ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document for the Oak Ridge National 

Laboratory -- Occupational Internal Dose

Document Number:

ORAUT-TKBS-0012-5 

Effective Date: 05/30/2006

Revision No.: 00 PC-1 

Controlled Copy No.: ________

Page 1 of 70

Subject Experts: Mike Bollenbacher, Kenny Fleming, Bill Joyce

Document Owner

Approval: Signature on File Date: 06/18/2004

Robert E. Burns, Jr., TBD Team Leader

Approval: Signature on File Date: 06/29/2004

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 06/28/2004

Richard E. Toohey, Project Director

Approval: Signature on File Date: 06/29/2004

James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

Record of Issue/Revisions ...................................................................................................................

Page

5

Acronyms and Abbreviations ...............................................................................................................6

5.0 Occupational Internal Dose.........................................................................................................8

5.1 Introduction................................................................................................................................. 8

5.1.1 Radionuclides of Concern ..................................................................................................9

5.1.2 Solubility Classes ............................................................................................................. 10

5.1.3 Route of Intake and Particle Sizes.................................................................................... 10

5.1.4 Bioassay Programs .......................................................................................................... 11

5.1.4.1 In Vitro Monitoring Program..................................................................................... 11

5.1.4.2 In Vivo Monitoring Program ..................................................................................... 11

5.1.5 Recordkeeping .................................................................................................................12

5.1.5.1 Division Codes......................................................................................................... 12

5.1.5.2 Radioanalytical Abbreviations and Codes................................................................ 12

5.1.5.3 Whole-Body Counting Results Codes...................................................................... 14

5.2 In Vitro Minimum Detectable Activities, Counting Methods, and Reporting Practices ................ 15

5.2.1 In Vitro Minimum Detectable Activities.............................................................................. 15

5.2.1.1 Minimum Detectable Activities between 1947 and 1989 .......................................... 15

5.2.1.2 Minimum Detectible Activities after 1989 ................................................................. 15

5.2.2 Counting Methods for In Vitro Samples ............................................................................ 15

5.2.2.1 Alpha Spectrometry ................................................................................................. 15

5.2.2.2 Liquid Scintillation.................................................................................................... 17

5.2.2.3 Zinc-Sulfide Scintillation........................................................................................... 17

5.2.2.4 Gamma Spectrometry.............................................................................................. 17
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5.2.2.5 Gas Flow Proportional Counter (Beta Counting)...................................................... 18

5.2.3 Notes on Measurements of Alpha Emitters....................................................................... 18

5.2.3.1 Trivalent Alpha Actinides ......................................................................................... 18

5.2.3.2 Plutonium ................................................................................................................ 18

5.2.3.3 Environmental Uranium ........................................................................................... 18

5.3 In Vivo Minimum Detectable Activities, Counting Methods, and Reporting Practices ................ 19

5.3.1 Shielded Counting Room.................................................................................................. 19

5.3.2 Detectors, Geometries, and Techniques .......................................................................... 20

5.3.3 Cesium Counting Artifact.................................................................................................. 25

5.3.4 Minimum Detectable Activity ............................................................................................ 26

5.4 Using Air Monitoring Data To Assign Best Estimates of Dose...................................................28

References ........................................................................................................................................30

ATTACHMENT 5A, CALCULATION OF MDAs FROM HISTORICAL DATA...................................... 33

5A.1 Recovery of Historical Data.......................................................................................................34

5A.1.1 Data Description............................................................................................................... 34

5A.1.2 Data Classification............................................................................................................ 34

5A.1.3 Synopsis of Recovered Data ............................................................................................ 35

5A.2 Calculation Method for Sample MDAs.......................................................................................35

5A.2.1 Standard Deviation of Background Counts....................................................................... 36

5A.2.2 Background Count Time................................................................................................... 37

5A.2.3 Instrument Efficiency ........................................................................................................ 37

5A.2.4 Yield................................................................................................................................. 37

5A.2.5 Volume of Aliquot Analyzed.............................................................................................. 37

5A.2.6 Volume of 24-Hour Sample .............................................................................................. 37

5A.3 Calculation Method for MDAs for Various Periods Between 1945 and 1988 ............................. 37

5A.4 Additional Considerations .........................................................................................................38

ATTACHMENT 5B, ASSESSMENT OF INTERNAL DOSE AT ORNL: 1944 to 1947........................ 67
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 3 of 70

LIST OF TABLES

Table

5-1 Radioanaytical results between 1947 and 1988 .....................................................................

Page

10

5-2 Solubility classifications used by ORNL for re-evaluation of historical bioassay 

results ....................................................................................................................................10

5-3 Routine bioassay monitoring methods and frequencies from 1973 to 1982............................ 11

5-4 Historical division codes.........................................................................................................12

5-5 New division codes ................................................................................................................13

5-6 EDP urinalysis abbreviations and codes ................................................................................14

5-7 Fecal analysis codes..............................................................................................................14

5-8 Result codes for whole-body counting....................................................................................14

5-9 Recommended in vitro MDAs (dpm/24-hr sample) for radionuclides from 1943 to 

1989.......................................................................................................................................16

5-10 MDAs for in vitro samples after 1989......................................................................................17

5-11 Construction history of the counting room ..............................................................................20

5-12 1965 selection criteria for whole-body counting......................................................................21

5-13 Qualitative information concerning radioactive material detected utilizing the 

WBC facility from 1966 to 1983..............................................................................................22

5-14 Qualitative information concerning radioactive material detected utilizing the 

WBC facility through 1978......................................................................................................23

5-15 Maximum measured in vivo activity (nCi) from 1961 to 1966..................................................24

5-16 Mean body burdens of 137Cs from fallout in United States ......................................................26

5-17 MDA values for six-detector HPGe array................................................................................26

5-18 Current MDA values for ORNL scanning bed counter (NaI/Ge).............................................. 27

5-19 Current MDA values for ORNL germanium thyroid counter....................................................27

5-20 Current MDA values for ORNL fixed bed counter...................................................................27

5-21 Frequency of in vivo monitoring..............................................................................................28

5A-1 Data used to determine historical MDAs for radionuclides in urine and feces......................... 36

5A-2 MDAs (dpm/24-hr sample) calculated from recovered data for radionuclides ......................... 39
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LIST OF FIGURES

Figure

5-1 Current MDA values for ORNL germanium lung counter versus chest-wall 

thickness................................................................................................................................

Page

27

5-2 Current MDA values for ORNL phoswich lung counter versus chest-wall 

thickness for 241Am and 238Pu................................................................................................. 27

5A-1 MDA data for americium-241 in urine.....................................................................................40

5A-2 MDA data for curium-244 in urine...........................................................................................41

5A-3 MDA data for Cs-137 in urine................................................................................................. 43

5A-4 MDA data for gross alpha in urine..........................................................................................44

5A-5 MDA data for gross beta in urine............................................................................................46

5A-6 MDA data for H-3 in urine.......................................................................................................47

5A-7 MDA data for Pm-147 in urine................................................................................................49

5A-8 MDA data for Pu-238 in urine................................................................................................. 50

5A-9 MDA data for Pu-239 in urine................................................................................................. 52

5A-10 MDA data for Pu-241 in urine................................................................................................. 53

5A-11 MDA data for rare earths in urine ...........................................................................................55

5A-12 MDA data for Sr-90 in urine ....................................................................................................56

5A-13 MDA data for U-233 in urine...................................................................................................58

5A-14 MDA data for U-238 in urine...................................................................................................59

5A-15 MDA data for Am-241 in feces ...............................................................................................61

5A-16 MDA data for gross alpha in feces .........................................................................................62

5A-17 MDA data for Pu-239 in feces ................................................................................................64

5A-18 MDA data for Th-232 in feces ................................................................................................65

5B-1 Plutonium in urine samples from 1945 ...................................................................................68

5B-2 Plutonium urine sample data through April 18, 1945 ..............................................................68

5B-3 Plutonium urine sample data for April 25-May 30, 1945 .........................................................69

5B-4 Alpha air monitoring results for 1944-1947.............................................................................69
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE REV. NO. DESCRIPTION

06/29/2004 06/29/2004 00 New technical basis document for the Oak 

Ridge National Laboratory – Occupational 

Internal Dose. First approved issue. Initiated 

by Robert E. Burns, Jr.

06/29/2004 05/30/2006 00 PC-1 Page change revision to update required 

language in the Introduction on page 8 in 

Section 5.1. Adds Section 5.4 on pages 28 and 

29 on using air monitoring data to assign best 

estimates of internal dose for unmonitored 

ORNL workers during the 1944-1947 period. 

Makes reference changes on pages 30 and 32 

in the Reference Section. Adds Attachment 5 

B, on pages 67, 68, 69, and 70, which gives the 

technical justification for this protocol. No 

changes occurred as a result of internal formal 

review. Incorporates NIOSH formal review 

comments. No sections were deleted. This 

revision results in no change to the assigned 

dose and no PER is required. Training 

required: As determined by the Task Manager. 

Initiated by Elizabeth M. Brackett and Thomas 

R. La Bone.

Approval: 

John M. Byrne, TBD Team Leader

Signature on File 05/22/2006 

John M. Byrne, Task 3 Manager

Signature on File 05/22/2006

Edward F. Maher, Task 5 Manager

Keith McCartney Signature on File for 05/22/2006

Kate Kimpan, Project Director

Signature on File 05/22/2006

James W. Neton, Associate Director for Science

Signature on File 05/30/2006
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 6 of 70

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AMAD activity median aerodynamic diameter

cm centimeter

cm2 square centimeter

cnts counts

cnts/dis counts per disintegration

cnts/min counts per minute

dia diameter

dpm disintegrations per minute

EDP Electronic Data Processing

EEOICPA Energy Employee Occupational Illness Compensation Program Act

ft feet

g gram

g/24-hr grams per 24-hour sample (fecal)

HPGe hyper-pure, intrinsic germanium 

hr hour

ID identification

ICRP International Commission on Radiological Protection

in. inch

IVGS in vivo gamma-ray spectrometer

keV kilovolt-electron, 1,000 electron volts

L liter

m meter

μm micrometer

MDA minimum detectable activity

ml milliliter

ml/24-hr milliliter per 24-hour sample (urine)

min minute

MMES Martin Marrieta Energy Systems 

mo month

MPBB maximum permissible body burden

MPOB maximum permissible organ burden

N population size

NaI sodium iodide

NaI-CsI sodium iodide – cesium iodide phoswich detector

NaI(Tl) thallium drifted sodium iodide

NaI/Ge sodium iodide / germanium

nCi nanocurie
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NCRP National Council on Radiation Protection and Measurement

NIOSH National Institute for Occupational Safety and Health

ORNL Oak Ridge National Laboratory

pCi picocurie

SBC scanning bed counter

TBD technical basis document

U.S.C. United States Code

WBC Whole-Body Counting (Facility)

wk week
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5.0 OCCUPATIONAL INTERNAL DOSE

5.1 INTRODUCTION

Technical basis documents (TBDs) and Site Profile documents are general working documents that 

provide guidance concerning the preparation of dose reconstructions at particular sites or categories 

of sites. They will be revised in the event additional relevant information is obtained about the 

affected site(s). These documents may be used to assist NIOSH in the completion of the individual 

work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 

U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation 

exposures at the facility in its dose reconstructions for employees at DOE facilities, including radiation 

exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and external 

dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts are made 

to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC.

 
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 9 of 70

The Oak Ridge National Laboratory (ORNL) began operations in early 1943. The startup of the 

Graphite Reactor and plutonium separation activities in late 1943 introduced the potential for 

personnel exposures from intakes of radioactive material. Laboratory operations involving radioactive 

materials increased over subsequent years as ORNL expanded its roles in radionuclide production 

and development of chemical separations processes. 

Development of methods and techniques for internal monitoring (bioassay) was one of the many 

priorities at ORNL in its early years of operation, because such methods simply did not exist. 

Although ORNL used air sampling and radiological contamination monitoring programs as qualitative 

indicators of internal exposure, urinalyses for various internal contaminants did not begin at the site 

until about 1947. (A limited number of in vivo measurements appear to have begun at ORNL in 

1959.) ORNL maintained early tolerance levels for airborne contamination based on “product” (i.e., 

239Pu) concentrations in the air (Cox 1944). [These early tolerance levels for alpha and beta-gamma 

contaminants were 3E-11 and 1E-07 μCi/cc respectively during the mid 1940s.] In addition, the 

Laboratory later established tolerance levels for materials such as 131I and noble gases.

As written, this technical basis document (TBD) is applicable for the period 1947 to the present. 

TBDs and Site Profile Documents are general working documents that provide guidance concerning 

the preparation of dose reconstructions at particular sites or categories of sites. They will be revised 

in the event additional relevant information is obtained about the affected site(s). These documents 

may be used to assist NIOSH in the completion of the individual work required for each dose 

reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy facility” as defined in the Energy Employee 

Occupational Illness Compensation Program Act (EEOICPA) of 2000 (42 U.S.C. § 7384l (5) and (12)).

5.1.1

Because of the many diverse processes and experiments that took place there, a complete list of 

radionuclides encountered at ORNL would be difficult to assemble. Radionuclides likely to produce a 

measureable internal dose include uranium, activation products, fission products, and transuranics. 

The earliest urine sample results that were provided by ORNL were for the isotopes 239Pu and 90Sr. 

[The electronic data that were provided by ORNL for use in estimating isotopic MDAs came from a 

project performed in the early to mid-1990s to convert hardcopy data over to a dBase IV database. 

Funding ran out on the conversion project in the mid-1990s and the entire set of site data was never 

completely converted, but a significant number of results were made available for our use. Table 5-1 

lists radionuclides included in in vitro bioassay results provided by ORNL for the period from 1947 to 

1988. The results were provided for estimating minimum detectable activities (MDAs) for various 

analyses. The source of these data was an electronic database created by ORNL in the early 1990s 

from hard-copy bioassay records. The sample size values in Table 5-1 are the numbers of analyses 

for that nuclide included in the data ORNL provided. These values do not reflect the total number of 

in vitro bioassays performed by ORNL in this period, because not all hard-copy records are in the 

database.

Radionuclides of Concern

Isotope-specific analyses for in vitro samples did not become routine until 1989. Prior to that time, 

chemical methods were used to separate radioelements as well as practicable, and the materials 

were assayed in terms of total activity. The activity measured would later be assigned to a 

predominate nuclide. Thus, a result from the early years might indicate 90Sr, when in reality it includes 

89Sr. The same is true for early plutonium results and results for transuranic materials. Thus, 

"associated" radionuclides are inherently included in such results. Process knowledge of 

radionuclides present in various work areas was used to assign nuclides to sample results. The 
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Table 5-1. Radioanaytical results between 1947 and 1988. 

Nuclide Sample size Nuclide Sample size Nuclide Sample size

Am-241 5,670 I-131 41 S-35 10

Am-243 12 Mn-54 2 Sb-125 1

As-74 7 Mo-99 1 Sm-151 11

BG 2 Na-24 3 Sr-85 1

Bk-249 14 Nb-95 3 Sr-89 37

Br-82 2 Np-237 55 Sr-90 12,893

Br-83 2 P-32 166 Tc-99 20

C-14 11 Pa-231 55 Th-232 1,125

Ca-45 4 Pa-233 16 Tl-201 1

Ce-144 37 Pa-234 1 Tl-204 1

Cf-249 3 Pb-210 2 Tm-170 6

Cf-252 14 Pm-147 80 U-232 1

Cl-36 1 Po-210 66 U-233 829

Cm-242 12 Pu-238 65 U-235 3

Cm-244 299 Pu-239 15,476 U-238 11,434

Co-60 83 Pu-241 112 U-239 11

Cs-134 1 Pu-242 41 Y-88 5

Cs-137 3,561 Ra-226 333 Y-90 31

Fe-59 9 Ra-228 1 Zn-65 7

Gross alpha 4,875 Rare earths 1,098 Zr-95 20

Gross beta 324 Ru-103 1

H-3 2,070 Ru-106 65

present ORNL internal dosimetry program uses a limited number of radionuclides for screening 

purposes. Positive results are followed up with additional bioassays. 

5.1.2

Internal dosimetrists at ORNL provided a list of assumed intake modes and clearance class 

information for 73 radionuclides. These assumptions, given in Table 5-2, are those used by ORNL 

when re-evaluating historical bioassay results. They are provided as a reference for dose 

reconstructors, with the caveat the assumed solubility classes are not known to be based on any 

specific studies. In general, they merely reflect conservative choices for dose assessment in terms of 

committed effective dose equivalent. Dose reconstructors therefore should not assume that these 

classes, which are in terms of the system promulgated in (ICRP 1966), represent claimant-favorable 

choices for tissue-specific dose evaluations. Dose reconstructors should instead make a claimantfavorable choice for the radionuclide and tissue(s) of interest using the system of solubility classes 

described in ICRP Publication 66 (ICRP 1994). 

Solubility Classes 

5.1.3

Unless additional information is provided, it should be assumed in all cases that the route of intake for 

internally deposited radionuclides was via inhalation of 5 μm Activity Median Aerodynamic Diameter 

(AMAD)-sized particles. 

Route of Intake and Particle Sizes

Table 5-2. Solubility classifications used by ORNL for re-evaluation of historical bioassay results. 

Material 

class

Intake 

mode Radionuclides

V (Very soluble) Ingestion H-3

L (Labeled organic) Inhalation C-14

D (Days) Inhalation Na-22, P-32, P-33, S-35, Rb-86, Sr-85, Sr-89, Sr-90, I-125, I-129, I-131, 

Ba-133, Cs-134, Cs-137, Eu-152, Eu-154, W-188

W (Weeks) Inhalation Ca-45, Cr-51, Mn-54, Fe-55, Fe-59, Ni-63, Ge-68, Sc-75, Tc-99, Gd-153, 

Hg-203, Bi-207, Po-210, Ra-226, Np-237, Am-241, Am-243, Cm-242, 

Cm-244, Cm-248
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Y (Years) Inhalation Sc-46, Co-57, Co-58, Co-60, Cu-64, Zn-65, Cu-67, Y-88, Y-90, Pd-103, 

Ru-106, Cd-109, Ag-110m, Pm-147, Ir-192, Os-191, Th-228, Th-229, 

Th-230, Th-232, Pa-231, U-232, U-233, U-234, U-235, U-236, U-238, 

Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Cf-249, Cf-252

5.1.4

5.1.4.1 In Vitro Monitoring Program

Bioassay Programs

ORNL has collected urine and fecal samples from individuals suspected of potential intakes from 

1947 to the present. Urine samples have been and still are the preferred method. While fecal 

samples can provide good supplementary information to determine when an intake occurred, the 

chemical solubility of the material, and particle size; there typically is more variation associated with 

these samples than with urine samples. When fecal samples are obtained with urine samples after a 

known intake, the results can be used to better understand the intake parameters and provide a more 

accurate estimate of intake. 

Urine samples were collected in the early years of the bioassay program based on the area health 

physicist’s knowledge of field conditions (e.g., known spills/incidents, air and contamination sample 

results, etc.). This practice of scheduling did not utilize a specified sampling frequency (Auxier 2004; 

Henley 2004). A 1961 procedure manual (ORNL 1961) references procedures and practices 

governing the health physics program at that time, including internal and external exposure 

monitoring. Although referenced, the procedure detailing internal dosimetry was not among the 

documents available for review.

A similar manual of health physics procedures and practices from 1982 lists a detailed set of 

requirements for graduated routine sampling, depending on the frequency and extent of a potential 

exposure (ORNL 1982). Table 5-3 lists ORNL published routine bioassay monitoring methods and 

frequencies from 1973 through 1982. NOTE: Historical information of the in vivo monitoring program 

(see Section 5.3) indicates that the whole-body counting frequencies listed in Table 5-3 were not 

consistently followed. Discussions with previous site personnel indicate that no formal counting 

frequency was used at ORNL until the later 1980s (Berger 2004). 

Section 5.2 discusses the in vitro monitoring program and the data it produced. 

5.1.4.2 In Vivo Monitoring Program

ORNL has collected whole-body, lung, and wound counting data for employees since 1959. For the 

most part, in vivo counting was used until the late 1980s to confirm potential intakes from known 

incidents or identified by the in vitro monitoring program. Though Morgan et al. (1965) indicated that 

routine in vivo monitoring for all site radiological workers began in 1965, it appears that a formal 

Table 5-3. Routine bioassay monitoring methods and frequencies from 1973 to 1982a (ORNL 

1982)

Radioactive material

Routine sampling categories

I II III

Pu-241 and alpha emitters 

other than uranium

Urine 3-4 wk

Whole-body 3-4 mo

Urine 6-13 wk

Whole-body 6-13 mo

Urine 6-12 mo

Whole-body 6-12 mo

Sr-90 and uranium Urine 4-6 wk Urine 4-13 wk Urine 6-12 wk

H-3 Urine each wk Urine each mo Urine each qtr

I-131 Whole-body each wk Whole-body each qtr Urine each qtr

Co-60 and Cs-137 Whole-body 3 mo Whole-body 6 mo Whole-body each 6 to 12 mo

All others Consult w/Internal Dose Group

a. Frequency of sampling should be in accord with employee's potential for exposure as determined by health 

physics surveyors. The following is a guide:
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 12 of 70

Category I Persons actively involved in operations or processes containing quantities of radioactive material, 

and when there is some evidence of contamination (i.e., positive results from smear and air 

samples).

Category II Employees working with relatively small quantities of materials that are confined or when there is 

no evidence of contamination or activity.

Category III Employees working with radioactive material or in the vicinity of material when there is no known 

exposure but some potential for exposure.

program did not begin until the mid- to late 1980s. Section 5.3 discusses the in vivo monitoring 

program and the data it produced. 

5.1.5

ORNL used various formats on paper records to record bioassay results through most of the period of 

interest. In addition, ORNL has entered much historical in vitro monitoring data into a database. 

These data were used to estimate MDAs as described in this document (see Attachment 5A). 

Discussions with personnel responsible for radiological records (Dixon 2004) indicated that hard-copy 

data are maintained at ORNL and that upon formal request these records are made available to the 

requester. These records will be used by dose reconstructors in estimating intakes. 

Recordkeeping

It is not practical for this document to provide examples of all record formats used in the past six 

decades; MMES (1995) contains a guide to historical record formats. Most bioassay records share

the following information: name, badge or other ID number, division code, health physics area, date, 

analysis code, results (in disintegrations per minute per 24-hr sample), and a reason for the analysis. 

A review of ORNL claim files indicated that many bioassay forms had slight revisions throughout their 

use, but that the information they maintain is similar from one revision to the next and should be easily 

interpreted. 

5.1.5.1 Division Codes

The division code sometimes provides information on individual locations and job assignments. Table 

5-4 lists historical division codes, and Table 5-5 lists more recent codes. 

5.1.5.2 Radioanalytical Abbreviations and Codes

Paper records contain abbreviation codes for recording analytes. These codes were often based on 

the initials of the analysis or the isotopic abbreviation of the radionuclide. Tables 5-6 and 5-7 list the 

codes for radionuclides measured in urine and fecal samples, respectively (Mani 1983) (MMES 1995). 

Table 5-4. Historical division codes (MMES 1995).

Codes DIVNUM Departments Division name

AC 01 3390 Analytical Chemistry

BI 02 4455 Biology

CH 04 Chemistry

CT 03 3370 Chemical Technology

DI 20 3200 Directors

EC, GE 38 3060 Gen. Engr. and Construc.

ED 06 3480 Education

EL 07 3320 Electronuclear

FM 22 Finance and Materials

HE 23 3090 Health

HP 08 3810, 4193, 3193, 3490 Health Physics

IC 09 3341, 3075 Instrumentation and Controls

IE 24 Inspection Engineering

IS 25 3369, 3650, 4362, 3360 Isotopes

LP 26 3094 Laboratory Protection

MA 10 3166, 3516, 3152 Mathematics

MC 11 3470 Metals and Ceramics

MET Metallurgy

NP 12 3410 Neutron Physics

OP 28 3639 Operations

PE 21 3016, 3078, 3003, 3004, 3062 Plant and Equipment
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PH 13 3405 Physics

PI 30 3173 Public Information

PR 29 3107, 3141 Personnel

RC 14 3430, 4430 Reactor Chemistry

RE 16 4435, 3435 Reactor 

RP 17 Research Participation

SS 18 3475 Solid State

TH 19 4460 Thermonuclear

TI 31 3072, 3148 Technical Information

Table 5-5. New division codes (MMES 1995)

Code DIVNUM Division

AC 01 Analytical Chemistry

AS 70 Administrative Services

AT 07 Applied Technology

BI 02 Biology

BS 95 Business Systems

CH 04 Chemistry

CM 20 Central Management

CS 63 Computing and Telecom

CT 03 Chemical Technology

EA 93 ESA

EC 35 Environmental Compliance

EH 36 Env. & Health Prot.

EN 15 Energy

EP 12 Eng. Physics & Math

ER 29 Employee Relations

ES 42 Env. Sciences

ET 16 Eng. Technology

EX 90 Executive Offices

FE 19 Fusion Energy

FM 37 Finance and Materials

FR 14 Fuel Recycle

GE 69 ORNL Engineering

GR 71 Graphics

HE 23 Health

Code DIVNUM Division

HS 08 Health & Safety RSRH

IC 09 Instrmt & Controls

IF Isotopes

IR 43 Info. Resource Org.

IS 72 Isotopes Division

LP 26 Lab Prot.

MC 11 Metals and Ceramics

OC 62 Controller Office

OP 27, 32 Operations

OS 22 Operational Safety

PC 64 Procurement

PE 21 Plant & Equipment

PH 13 Physics

PU 73 Publications

QA 24 Quality

RE 87 Env. Restoration

RP 14 Robotics

RR 06 Research Reactors

RU Rust

SS 18 Solid State

TR 60 Treasurer Office

VI Visitor

WM 27 Waste Management
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 14 of 70

Table 5-6. EDP urinalysis abbreviations and codes. 

Alphanumeric code Numeric code

CM0 244Cm 000 Other

CO0 60Co 001 35S

CS0 Cesium B (137Cs) 002 60Co

CS7 137Cs 003 210Pb

EU 154Eu 004 24Na

FP0,FP Fission products (137Cs) 005 95Zr/95Nb

FU0 Total rare earths () 006 99Tc

GA0,GA Gross alpha (239Pu) 007 74As

GB0,GB Gross beta (90Sr) 008 82Br/83Br

GD0 153Gd 009 59Fe

GG0 Gross gamma (137Cs) 010 54Mn

GU0 Gross alpha (239Pu) 011 131I

HY3 3H 012 132Cs

I25 125I 013 Gross beta

I31 131I 014 140Ba

NP0 237Np 015 125Sb

PA0 230Pa alpha 016 204Tl

Alphanumeric code Numeric code

PA3 233Pa beta 017 237Np

PH2 32P 018 110mAg

PM7 147Pm

PO0 210Po

PU0,PU 239Pu alpha 

PU1 241Pu

PU9 239Pu

RA0 226Ra

RU6 106Ru

SR0 90Sr

SR5 85Sr

SR9 89Sr

TA0 180Ta

TH 232Th

TP0 Trans plutonium alpha (241Am)

TRE Total rare earths ()

URO 234U alpha 

Table 5-7. Fecal analysis codes. 

Alphanumeric code

GF0 239Pu gross alpha (includes Th)

PF0 239Pu

RF0 Rare earths ()

SF0 90Sr

SF9 89Sr

TF0 Trans plutonium (241Am)

UF0 234U

OF0 Other

5.1.5.3 Whole-Body Counting Results Codes

Table 5-8 lists codes used on some older cards and forms for documenting whole-body counting 

results. Many hardcopy records have been consolidated into individual personal records folders. 

However, this compilation is incomplete, with records for only employees with last names beginning 

with A through G. 

Table 5-8. Result codes for whole-body counting.

Before 1971 1971-1978

Code Description* Code Description*

1 Normal human spectrum 0 = <15% MPOB

2 Less than 10% MPBB 2 = <25% MPOB

3 Less than 25% MPBB 4 = <50% MPOB

4 Less than 50% MPBB 6 = <100% MPOB

5 Greater than 50% MPBB 8 >100% MPOB

N Insignificant & indeterminate

S Significant & indeterminate

*MPBB = maximum permissible body burden; MPOB = maximum permissible organ burden. 
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 15 of 70

5.2 IN VITRO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PRACTICES

5.2.1

5.2.1.1 Minimum Detectable Activities between 1947 and 1989 

In Vitro Minimum Detectable Activities 

Table 5-9 lists historical MDAs for radionuclides of concern in urine and feces. These MDAs were 

calculated from analytical records recovered from the in vitro sample databases supporting the ORNL 

historic workforce dose assessment project (MMES 1995). Attachment 5A contains details of data 

recovery and additional information on specific radionuclides. Blank MDA entries in the early years in 

Table 5-9 indicate that no analytical results for that radionuclide were recovered from that year. This 

could be because the records have not been located, or because the analysis was not performed.

Recovered data show that analytical MDAs tend to remain fairly consistent for a number of years. 

Abrupt changes in MDAs can be identified for groups of radionuclides during specific years. Following 

these changes, the MDAs remain generally consistent in subsequent years. This “step-wise” pattern 

allowed MDAs from several years to be grouped to obtain a single, representative MDA. Table 5A-2 

in Attachment 5A provides annual averages for periods of time where bioassay data were available. 

The ORNL database contains values for isotopic activities during times when isotope-specific 

analyses were not possible or routinely performed. In earlier years, the element of concern was 

extracted chemically from the biological sample and the total radioactivity of the element in the extract 

was measured. At some point after the extraction and sample count, the total sample activity was 

attributed to a specific radionuclide. Many of the isotopic assignments were based on process 

knowledge. 

To reflect the performance of the instrumentation and analytical methods more accurately during the 

period prior to 1990, Table 5-9 assigns MDAs for some isotopes to their corresponding radioelement 

rather than the specific radionuclide. For example, routine separation of the alpha emitters 238Pu and 

239Pu did not occur until alpha spectrometric analyses became routine in 1989, but the recovered 

database reports both separately. They have been combined in this document as “Plutonium.” 

NOTE: 241Pu is reported separately because it is a beta emitter that can be assayed separately from 

the alpha-emitting isotopes of plutonium. 

5.2.1.2 Minimum Detectible Activities after 1989

MDA values for present (since 1989) samples are available with each sample. Table 5-10 lists 

typical, current MDAs for radionuclides of concern in urine and feces. 

5.2.2

Several counting methods have historically been available at ORNL for determining radioactivity in in 

vitro samples. The following sections discuss alpha spectrometry, liquid scintillation, zinc sulfide 

scintillation, gamma spectrometry, and beta counting using a gas flow proportional counter. 

Counting Methods for In Vitro Samples

5.2.2.1 Alpha Spectrometry 

ORNL uses alpha spectrometry in the analysis of nuclides that decay primarily by alpha emission, 

with only very low-energy photons or none at all. A tracer is added to the bioassay sample before 

analysis begins to determine the chemical yield of the process. The radioelements are chemically 
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 16 of 70

1 Table 5-9. Recommended in vitro MDAs (dpm/24-hr sample) for radionuclides from 1947 to 1989. 

Year

Urine MDAs (dpm/24h sample) Fecal MDAs (dpm/24h sample)

Gross 

alpha

Gross 

beta

Am241Cm244Cs137 H-3I131Np237Pm147 Plutonium PoloniumPu241Ra226Rare 

earths

Ru106Sr-89 + 

Sr-90 Uranium

Gross

alpha

Am241Cm244 PlutoniumTh232

1943

1944

1945

1946

1947 0.38

1948 0.38 1.4

1949 0.38 1.4

1950 33 0.21 34 1.4 0.53

1951 0.33 33 0.21 129 0.30 34 6.3 0.53

1952 0.26 0.33 33 0.21 31 129 0.30 34 6.3 0.53

1953 0.26 0.33 33 0.21 58 31 129 0.30 34 6.3 0.53

1954 0.26 1,135 0.33 33 0.21 58 1.3 129 0.30 34 6.3 0.53

1955 0.26 1,135 0.33 197 0.21 58 1.3 129 0.30 34 6.3 0.53

1956 0.26 1,135 0.33 197 0.21 58 1.3 129 0.30 34 6.3 0.64 0.53

1957 0.26 1,135 0.33 197 1.1 0.21 58 1.3 129 0.30 34 6.3 0.64 0.34 0.53

1958 0.26 1,135 0.33 197 1.1 0.21 58 1.3 129 0.30 34 6.3 0.64 0.34 0.53

1959 0.26 1,135 0.33 197 1.1 396 0.21 58 1.3 129 0.30 34 6.3 0.64 0.34 0.53

1960 0.26 1,135 0.33 197 1.1 396 0.21 58 1.3 129 53 34 6.3 0.64 0.34 0.53

1961 0.26 288 0.33 197 320,282 1.1 396 0.21 17 1.3 129 53 6.3 6.3 0.64 0.34 0.53

1962 0.26 288 0.33 197 320,282 1.1 396 0.21 17 1.3 129 53 6.3 6.3 0.64 0.34 0.53

1963 0.26 288 0.33 197 320,282 1.1 396 0.21 17 1.3 129 53 6.3 6.3 0.64 0.34 0.53

1964 0.26 288 0.33 0.22 197 320,282 0.09 396 0.21 17 1.3 129 53 6.3 1.1 0.64 0.34 0.53

1965 0.26 288 0.33 0.22 197 320,282 10,358 0.09 396 0.21 17 1.3 129 53 6.3 1.1 0.64 0.46 1.05 0.34 0.53

1966 0.26 288 0.33 0.22 197 320,282 10,358 0.09 396 0.21 17 1.3 129 53 6.3 1.1 0.64 0.46 1.05 0.34 0.53

1967 0.26 288 0.33 0.22 197 320,282 10,358 0.09 396 0.21 17 1.3 129 53 6.3 1.1 0.64 0.46 1.05 0.34 0.53

1968 0.09 39.8 0.08 0.10 51 55,681 10,358 0.09 396 0.06 17 9.2 1.3 129 53 6.3 0.09 0.64 0.46 1.05 0.34 0.53

1969 0.09 39.8 0.08 0.10 51 55,681 10,358 0.09 396 0.06 17 9.2 1.3 129 53 6.3 0.09 0.64 0.46 1.05 0.34 0.53

1970 0.09 39.8 0.08 0.10 51 55,681 10,358 0.09 396 0.06 17 9.2 1.3 129 53 6.3 0.09 0.64 0.46 1.05 0.34 0.53

1971 0.09 39.8 0.08 0.10 51 55,681 10,358 0.09 396 0.06 17 9.2 0.07 5.4 53 4.0 0.09 0.64 0.46 1.05 0.34 0.53

1972 0.09 39.8 0.08 0.10 51 55,681 10,358 0.09 396 0.06 17 9.2 0.07 5.4 53 4.0 0.09 0.64 0.46 1.05 0.34 0.53

1973 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 17 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1974 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 17 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1975 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1976 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1977 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1978 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1979 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1980 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1981 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1982 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1983 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1984 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1985 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1986 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1987 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1988 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53

1989 0.09 39.8 0.05 0.10 51 55,681 10,358 0.09 396 0.04 0.52 9.2 0.07 5.4 53 4.0 0.06 0.64 0.46 1.05 0.34 0.53
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 17 of 70

Table 5-10. MDAs for in vitro samples 

after 1989 (McLaughlin 2002). 

Isotope

MDA 

(dpm/24-hr sample)

H-3 9,100

C-14 4,480

P-32 1

Sr-90 3

Tc-99 200

I-131 20

Np-237 0.02

Th-232 0.02

U-232 0.02

U-233 0.02

U-234 0.02

U-235 0.02

U-238 0.02

Pu-238 0.02

Pu-239 0.02

Am-241 0.02

Cm-244 0.02

Cf-252 0.02

Bk-249 26

separated from the sample and electrodeposited on stainless-steel disks. Plutonium and the other 

transuranics can be analyzed sequentially; uranium analysis requires a separate sample. 

5.2.2.2 Liquid Scintillation

Liquid scintillation is used for the analysis of low-energy, pure beta emitters, specifically 3

H and 14C. 

One milliliter of a urine sample is mixed with a scintillation cocktail for analysis. 

5.2.2.3 Zinc-Sulfide Scintillation

ORNL used zinc-sulfide scintillation counting to count alpha emitters like the trivalent alpha actinides 

(see Sections 5.2.3.1 and 5.2.3.2). 

5.2.2.4 Gamma Spectrometry

ORNL uses gamma spectrometry to identify and quantify radionuclides that emit photons with 

energies greater than 60 keV. A high-resolution, hyper-pure, intrinsic germanium (HPGe) detector 

with a beryllium end-window is used. Urine samples are placed in a 1-L MarineIli beaker, which is 

placed over the detector for counting. If the total sample volume is less than 1 L, distilled or deionized 

water is added to bring the volume up to 1 L. Fecal samples were counted directly in the sample 

counter for screening purposes, and they might have been ashed and placed into a 2-in. Petri dish for 

quantitative results. 
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 18 of 70

5.2.2.5 Gas Flow Proportional Counter (Beta Counting)

ORNL uses the gas flow proportional counter system for the analysis of strontium. Strontium is 

chemically separated from the sample and filtered onto a glass fiber filter. The filter is placed on a 

planchet for insertion into the counter. The counting system does not distinguish between beta 

energies, so the reported result is total strontium (89Sr plus 90Sr). 

5.2.3

5.2.3.1 Trivalent Alpha Actinides

Notes on Measurements of Alpha Emitters

Before 1989, ORNL did not perform radionuclide-specific analyses for americium, curium, and other 

high atomic number elements beyond plutonium. [These radionuclides were typically recorded on the 

HP Body Fluids Analysis Request cards as transplutonium (TPO or TPL).] Rather, the Laboratory 

separated trivalent alpha actinides as a group and analyzed by zinc-sulfide scintillation counting. 

Therefore, monitoring of transplutonium elements was unable to differentiate between such nuclides 

as 241Am and 244Cm. The default radionuclide to use with measurements involving trivalent alpha 

actinides would be 241Am. The detection sensitivity of the transplutonium analysis technique is not well 

documented for samples processed before 1985. However, laboratory records suggest that the 

transplutonium detection level was about 0.2 dpm through 1985. 

For analyses performed after 1985, an estimate of the sample-specific MDA is generally reported. 

Alpha spectroscopy analysis for transuranics in bioassay samples began in the early- to mid-1980s. 

Differentiation, by alpha energy separation, for isotopes such as 234U, 235U, 238U, 238Pu, 239Pu, 241Am, 

and 244Cm became possible.

5.2.3.2 Plutonium

Before 1989, ORNL did not routinely perform isotope-specific analyses for plutonium. Rather, the 

Laboratory separated plutonium as an element and analyzed it by zinc-sulfide scintillation counting. 

Therefore, the historic plutonium analysis technique was unable to differentiate between the alpha 

emitting isotopes 238Pu, 239Pu, and 240Pu. Site analytical personnel assert that in the early 1980s, 

positive total Pu measurements were recounted on the limited number of alpha spectrometers that 

were available at the site. Though many HP Body Fluids Analysis Request cards in the claims files 

gave results for the isotopes 238Pu and 239Pu, this was not consistently observed. The default isotope 

for positive, total Pu measurements should be 239Pu.

5.2.3.3 Environmental Uranium

The following paragraphs were taken entirely from the Oak Ridge National Laboratory Internal 

Dosimetry Program Technical Basis Document (McLaughlin 2002). They are included as an aid to 

dose reconstructors in the interpretation of uranium bioassay results for ORNL workers. 

Environmental levels of naturally-occurring uranium are found throughout eastern 

Tennessee. The environmental activity levels in the immediate area surrounding Oak 

Ridge are sufficiently high such that dietary intake of uranium is detectable with 24-

hour urine samples. A urinary uranium background study was conducted in the mid 

1990's using non-occupationally exposed employees to quantify the range of typical 

background uranium excretion. Based upon the results of that study, a discrimination 

level (set at the 99th percentile level) of 0.14 dpm/day was established for both U-234 

and U-238 to differentiate between environmental and occupational exposure to 
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 19 of 70

uranium. A value of 0.25 dpm/day is applied to total uranium results. Plots of the 

observered uranium excretion distributions for U-234, U-238, and total uranium are 

provided below.

An activity ratio of U-234 to U-238 of approximately 2 to 1 has been observed within 

the analyzed local potable water samples. Activity ratios in this range have been 

reported for various aquifers (Osmond). Though U-234 and U-238 should be in secular 

equilibrium within nature, the observed enrichment of the isotope U-234 is believed to 

be caused by several factors which include the direct transfer of U-238 decay products 

across a solid/liquid phase boundary by alpha recoil and differences in solubility 

between uranium decay chain members. Recognizing this trend, uranium bioassay 

results that are less than 0.2 dpm/day that do not exhibit a U-234 to U-238 activity ratio 

of 2:1 should be considered suspect and investigated.

5.3 IN VIVO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PRACTICES

5.3.1

The ORNL Whole-Body Counting (WBC) facility, sometimes referred to as the In Vivo Gamma-Ray 

Spectrometer or IVGS), is in Building 2008 in the northwest corner of the Main Plant area. Several 

Health Physics Division annual progress reports (Morgan et al. 1961, 1963, and 1966) indicate that 

the WBC facility began operation in June 1960 and another report (Brown 1971) indicates a May 1960 

date; a more recent document (Watts et al. 1995) indicates several counts occurred earlier. Thus, it is 

likely that the WBC facility began limited operations in 1959. 

Shielded Counting Room

The main counting room has inner dimensions of 10 ft by 10 ft by 10 ft. Its walls consist of four layers 

of pre-World War II steel with a total thickness of 14 in. Some documents called it the iron or big

room. The first recorded count in the WBC facility occurred on July 13, 1959, and, “was a background 

count [conducted] in the corner of the steel room with the door not in place and the roof incomplete”

(Watts et al. 1995). The room was completed on July 24, 1959 (with the exception of a 0.125-in. layer 

of “special low-radioactivity lead” added to all inner surfaces in 1960 to reduce background radiation 

levels) and was used for a time to conduct background studies of paint and interior samples and to 

count biological samples such as milk, grass, and cow thyroids. The first recorded lung count of an 

employee occurred on May 19, 1960. Almost immediately after (May 19 to 20, 1960), it was used to 

conduct lung counts of three employees involved in an onsite contamination incident. The original 

counting facility utilized a 4- by 4-in. NaI (Tl) crystal, but it is not clear what the counting geometry 

was. There is a photograph of an individual laying on a nylon-strapped, aluminum beach chair, but it 

was not clear whether that was the geometry initially employed in the facility. The Argonne Chair

counting geometry was set up for whole-body counting on February 27, 1961 (Morgan et al. 1961). 

The tubular steel chair was tilted so the individual’s body was in a V position with the detector placed 

approximately 50 cm directly over the hips (Morgan et al. 1961, p. 223).

Before construction of the WBC facility, health physicists realized that there could be problems with 

siting a low-level radiation counting facility in the plant environment, with radioactive effluents from 

operations being generated immediately adjacent to the facility. Thus, ORNL designed and installed a 

recirculating air treatment system that pumped air from inside the counting room through charcoal 

traps, cooling coils, and heaters to remove radon, odors, and excess moisture (Brown 1971; Morgan 

et al. 1961). A slight positive pressure was maintained in the counting room by using cylinders of 

“aged” breathing air to make up for leakage from the system. 
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 20 of 70

To improve the detection sensitivity, additional layers of material were applied to the floor in the WBC 

facility counting room over the years. Morgan et al. (1963) indicates that 0.04-in.-thick layers of tin 

and cadmium were laid over the interior surface of the lead, only on the floor, with a 0.01-in.-thick 

layer of copper over the tin and cadmium. The original vinyl tile was placed over the copper. These 

materials were installed to reduce background radiation emitted from the lead shield as a graduating 

shield to minimize the contribution of low-energy X-rays. In the mid-1990s, the vinyl tile surface was 

removed and a 0.03-in.-thick layer of stainless steel was placed directly over the other metal layers. 

Site personnel indicated that the stainless-steel layer was primarily for aesthetic purposes to cover the 

oxidized copper layer. With the addition of frictional surfaces for slip reduction, this was the latest 

form of the floor. Table 5-11 lists the construction history of the counting room.

Table 5-11. Construction history of the counting room. 

Installation date Material Thickness (in.)

1959 (entire facility) Pre-WW II steel 14 

1960 (entire facility) Special low-radioactivity lead 0.125

1963 (floor only) Tin 0.04

1963 (floor only) Cadmium 0.04

1963 (floor only) Copper 0.01

Mid-1990s Stainless steel 0.03

Subsequent improvements to the shielded counting room, detectors, and counting geometries are 

described below. The shielded room was used to conduct all in vivo measurements at ORNL until 

1992, when a Canberra scanning bed counter (SBC) was installed in Room 16 of Building 2008. 

A discussion with Berger (Berger 2003), who was responsible for the facility from the mid-1970s until 

the early 1980s, indicated that although the facility was in operation, a formal in vivo monitoring 

program was not in place until the late 1980s. Prior to that time, the WBC facility was used almost 

exclusively to either confirm known or suspected intakes of radioisotopes or for research purposes. 

Selection of individuals for counting was performed by field health physics personnel based on 

expected contaminants of concern and the probability of exposure until the early 1990s, when the 

selection of individuals for in vivo monitoring became the responsibility of the Internal Dosimetry staff. 

5.3.2

As described above, the initial in vivo counter at ORNL utilized a 4- by 4-in. NaI (Tl) crystal and a tilted 

chair counting arrangement (Morgan et al. 1961). In July 1961, counting activities in the WBC facility 

were suspended to modify the detector arrangement with the installation of an 8- by 4-in. NaI (Tl) 

crystal to replace the smaller detector to increase efficiency and reduce counting times. In 1962, 

several thin (5-in.-diameter by 0.0625-in.) NaI (Tl) crystals were installed in the facility to quantify lowenergy photons (e.g., mainly isotopes of plutonium). Also in that year, calibration studies were 

conducted using an arc-shaped geometry. This geometry was expected to result in less variation in 

counting efficiency than the chair geometry. (The arc geometry placed an individual laying in an 

arced position with the anterior portion of the body facing the detector at a distance of about 1 m, so 

each portion of the body was approximately the same distance from the detector (Mani 1983). 

Detectors, Geometries, and Techniques 

Morgan et al. (1963) indicated that the thin crystal detectors could see approximately 40 nCi of 238Pu, 

if there was a preexposure chest count. If there was no preexposure count, the detection capability 

was approximately 80 nCi. In 1963, a SBC replaced the chair geometry using the 8- by 4-in. NaI (Tl) 

crystal for whole-body counting. The bed and individual were moved under the stationary detector. 

This geometry was used to determine roughly the part of the body in which the gamma-emitting 

radioisotopes were located. The detection efficiency was approximately equivalent to that of the chair 
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 21 of 70

geometry (Morgan et al. 1963). Morgan et al. (1963) noted that the computer output provided a 

“gross” spectrum as well as having the ability to “strip the 40K and 137Cs background counts.” Several 

spectra observed during that period indicated a large portion of the “net” spectra with negative values. 

Morgan et al. (1964) noted that in March 1963 “written weekly reports of in vivo counting activities and 

results was begun.” (The research to generate this document did not find any of these reports). 

Morgan et al. (1965) reported that “Baseline counts on essentially every person with a potential for 

future exposure was completed in May 1965.” By the third week of May 1965, a prioritizing system for 

selecting individuals for whole-body counting was initiated. Table 5-12 describes this system. 

Although several located documents stated that baseline and specified monitoring frequencies were 

utilized to make in vivo measurements, Berger (2003) and McLaughlin (2004) indicated that a full in 

vivo monitoring program did not exist at ORNL until approximately 1994, when site internal 

dosimetrists became responsible for identifying personnel for counting. Prior to that time, the area 

health physicists selected individuals for in vivo monitoring. The area health physicists were 

responsible for determining radioisotopes to which the worker could have been exposed and counting 

frequencies. This led to inconsistent approaches to the selection of individuals for monitoring. 

Table 5-12. 1965 selection criteria for whole-body counting. 

Priority Selection criteria

1 Persons suspected of having sustained exposure

2 Persons being recounted as follow-up to initial elevated in vivo results

3 Persons who work directly with radioactive material (once every 3 mo)

4 Persons who work in areas where radioactive materials are handled, but do not work directly with 

the material (once every 6 mo)

5 New hires or other persons requiring a baseline count and a limited number of persons prior to 

termination

Morgan et al. (1966) indicated that the “job of determining which employees should be counted, the 

frequency with which they should be recounted, and preparing the necessary schedule cards and lists 

is requiring more time than had been anticipated.” For this reason, the WBC facility was still being 

used to verify if intakes had taken place rather than for routine monitoring of site personnel. A new 

Health Physics Division Report (Morgan et al. 1967) begun in 1966 included data on numbers of in 

vivo counts and basic statistics for in vivo monitoring. They included numbers of individuals whose 

results exceeded the U.S. Atomic Energy Commission (AEC) reporting level (50% of the permissible 

body burden averaged over the year) or some other specified amount. Table 5-13 summarizes this 

information. NOTE: The use of annual averages could be misleading in cases involving short-lived or 

rapidly cleared materials. Mani (1983, Table 15) provided similar data, but he reported all in vivo

counts that exceeded specified levels rather than annual averages. Table 5-14 lists this information. 

Table 5-15 summarizes the maximum activity measured through in vivo monitoring for various 

nuclides for the period from 1961 to 1966. 

The results for a given in vivo analysis contained the following information: name, badge number (or 

equivalent), division code, HP area, date, analysis code, and results code. Table 5-8 lists result 

codes for the period prior to 1978. The reason for the analysis was given. The ORNL Division codes 

are the same as those listed in Tables 5-4 and 5-5. [NOTE: The ICRP 2 methodology of reporting 

percentages of body or lung burdens was used at ORNL through 1983. AEC Manual Chapter 0502 

required “an evaluation of the radiation exposure status of an employee when monitoring techniques 

indicated that a body burden equaled or exceeded 50% of a maximum permissible limit.” A phone 

discussion with the current internal dosimetry staff (McLaughlin 2004) indicate that the staff that was 

in place in the 1960s through 1980s probably conducted dose assessments when necessary and that 
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 22 of 70

Table 5-13. Qualitative information concerning radioactive 

material detected utilizing the WBC facility from 1966 to 1983. 

Year

Number of persons exceeding the permissible body 

burden (based upon in vivo measurements)

1966a 0 persons exceeded 50% of permissible body burden

1967a 0 persons exceeded 50% of permissible body burden

1968a 0 persons exceeded 50% of permissible body burden

1969a 0 persons exceeded 50% of permissible body burden

1970b 0 persons exceeded 25% of permissible body burden

1971b 0 persons exceeded 25% of permissible body burden

1972b 0 persons exceeded 25% of permissible body burden

1973b 0 persons exceeded 25% of permissible body burden

1974b 2 (employees inhaled Cm-244 believed to range from 20 

(15 pCi) to 40% (30 pCi) of organ (lung) burden)

1975b 4 (1 - Cm-244 of ~50%, 1 – U-238 of <15%, 1 – Zn-65 

and Co-57 of <15%, and 1 – Co-60 of <15% of lung 

burden)

1976b 6 (all six appear to have had detectable amounts of Cm244 of <15% of the lung burden)

1977b 20 (8 – 239Pu and 241Am of <25%, 8 – 137Cs, 60Co, 153Gd, 

and 109Cd of <15%, and 4 – 125I and 123Te <25% of the 

lung burden)

1978b 6 (1 - Pu-241 and Am-241 of ~70% (lung), 1 – 131I-131 

minor amount that could not be quantified (thyroid), and 

4 – Co-60 of <25% (lung)

1979b 0 persons exceeded 25% of permissible lung burden

1980b 0 persons exceeded 10% of permissible lung burden

1981b 0 persons exceeded 10% of permissible lung burden

1982b 0 persons exceeded 10% of permissible lung burden

1983b 0 persons exceeded 10% of permissible lung burden

a. The AEC permissible level was 50% of the permissible body burden 

from 1966 through 1969.

b. From 1970, ORNL indicated in reports that in vivo measurements did 

not exceed specified amounts.

documentation would exist for these assessments. Since 1989, each in vivo measurement showing a 

positive result has been assessed, including making estimates of intake and dose. The ORNL 

Internal Dosimetry Program TBD (McLaughlin 2002) states that dose estimates of less than 1 mrem 

are reported as zero. The overwhelming number of in vivo measurements obtained at the facility 

indicate no elevated activity (McLaughlin 2004).] 

Mani (1983) indicated that a NaI-CsI phoswich detector was installed and operational in 1967, though 

Brown (1971) states “The several detectors used for measuring gamma radiation from the human 

body are all NaI (Tl) crystals.” The 1974 annual report of the Health Physics Division (Auxier et al. 

1975) stated that a phoswich detector was in place and apparently was installed between 1971 and 

1974, but no other information was located stating when the detector was placed in the WBC facility. 

A 9- by 9-in. NaI (Tl) crystal was installed in the WBC facility in 1974 (Auxier et al. 1975) and placed 

under the chest area of the counting bed. This detector was used for stationary chest counts. (A 

statement in Brown (1971) indicated that a distortion of the spectrum occurred if the large detector 

was used with the bed moving. Thus, it was not used for whole-body scans.) The whole-body scans 

performed during this period used the 8- by 4-in. detector at a distance of 12 in. from the bed surface 

and an elapsed counting time of 20 minutes. (If transuranic isotopes were suspected, the thin crystal 

assembly was used.) Stationary chest counts used either the 8- by 4-in. or thin crystal assembly in 
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 23 of 70

Table 5-14. Qualitative information concerning radioactive material detected utilizing the WBC facility through 1978 (Mani 1983). 

Count 

year

Insignificant 

and 

indeterminate

Significant 

and 

indeterminate

<15% 

of MPOB

Normal 

spectrum

<10/25% 

of MPOBa 

<25% 

of MPOB

<50% 

of MPOB

>50% 

of MPOB

<100% 

of MPOB

>100% 

of MPOB

Total 

counts

1962 59 9 3 71

1963 906 15 921

1964 1,485 6 20 13 31 1,555

1965 1,111 16 8 23 11 1,169

1966 632 11 12 28 13 696

1967 918 13 6 25 4 966

1968 815 135 12 6 968

1969 905 126 8 18 1,057

1970 499 14 2 1 2 518

1971 310 4 9 1 324

1972 255 4 259

1973 321 1 3 325

1974 246 2 8 3 6 7 6 278

1975 297 23 5 4 1 330

1976 215 6 29 3 1 254

1977 256 14 15 5 1 291

1978 257 12 30 3 3 1 306

Total 2,157 62 103 7,330 360 71 128 62 8 7 10,288

a. 10% before 1971 and 25% from 1971 to 1978.
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 24 of 70

Table 5-15. Maximum measured in vivo activity (nCi) from 1961 to 

1966. 

Isotope 1961 1962 1963 1964 1965 1966

24Na 140

46Sc 4

51Cr 320 42 Tracea

56Co 15

57Co 65

58Co 20 38 Trace

59Fe 40 15

60Co 80 13 2 220 5 12

64Co <5

65Zn 40 3 16 Trace

75Se 250

90Sr 40 4,500 800 626

95Zr/95Nb 30 11 14 47 18 39

106Ru/106Rh 131 47 200 30

125Sb 162 6 18

131I 12 200 28b 73 54 6

137Cs 440 360 570 310 104 92

144Ce/144Pr 30 <7 75 50

155Eu Trace

198Au

203Hg 800 0.0005 169 22

226Ra <5 <0.5c

233Pa 2,800

U (enriched) Trace Trace

a. Trace is not defined within any of the annual reports, but is provided here to 

give a qualitative “feel” for the amount present.

b. Thyroid.

c. micrograms.

contact with the chest depending on what contaminants were expected. When the 9- by 9-in. NaI (Tl) 

crystal was used the individual was positioned either face-up or –down on the bed and the bed was 

lowered just shy of contact with the face of the large detector. The time to conduct stationary chest 

counts typically ranged from 10 to 40 minutes and sometimes longer to ensure “better counting 

statistics” (Brown 1971). The arc counting geometry was described in (Brown 1971) with the 8- by 4-

in. detector positioned 6 ft above the center point of the counting bed. This geometry differs from that 

described in Mani’s assessment of ORNL internal dosimetry data (Mani 1983). The counting bed was 

placed on the floor with the ends supported to give the proper curvature. This geometry was intended 

for use only when the count rate was too high for the scanning or stationary chest counting methods. 

(No evidence of the arc counting geometry actually being used was noted in the references reviewed 

in the generation of this document.) 

In addition to the NaI (Tl) detectors described above, Brown (1971) indicated that three other 

detectors were in use within the counting room in 1971. A 3- by 3-in. NaI crystal was used to 

measure radiation emanated from specific organs (e.g., mainly for the thyroid, kidneys, liver, and 

spleen). Because of its smaller size, it could be placed directly over the organ of interest. Two other 

NaI crystals were employed in the WBC facility as wound probes: 2- by 2-in. and 1- by 0.03125 in., 

with the latter used to measure low-energy photons less than about 200 keV. 

The next bulk source of information describing instruments and procedures in the WBC facility was 

provided by Berger and Goans (1981), Berger and Lane (1981, 1982, 1984), and Berger (2004). 
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 25 of 70

Berger (2004) indicated that use of the phoswich detector began in late 1976 to early 1977. Until that 

time, the three thin NaI(Tl) crystal detector assembly was in use for detection of low-energy photons. 

The phoswich detector had the following dimensions: 5.25- by 0.0625-in. NaI - 2-in. CsI. Mani (1983) 

indicates that in 1978, an 80-cm2 HPGe counting array was in the WBC facility, but it was not fully 

operational until May 1980 (Berger and Lane 1981). The array consisted of six separate detectors in 

an aluminum block in a close-packed rectangular array (e.g., a six-die pattern) designed to cover one 

lung (Berger and Lane 1981). The phoswich and HPGe detectors were used together from the early 

1980s to obtain lung count data; they were positioned over the right and left lungs, respectively. 

Berger and Lane (1984) indicated that there were problems with getting all the HPGe detectors 

positioned close to the chest surface due to the construction of the aluminum block. Detectors on the 

chest surface were almost a factor of 3 more sensitive than those farther away from the chest. This 

lung count geometry was used until approximately 1987, when problems developed with the 

HPGe/liquid nitrogen feed system (Hembree 2004). At that time, the HPGe detector array was taken 

out of service, and the phoswich detector was the only device used for routine chest counts until 

1994. 

In 1992, a Nuclear Data SBC was put in service, primarily to measure mixed fission/activation 

products ranging in energy between 100 and 2000 keV, and is in use today. This counter is in Room 

16 of Building 2008, and is the only in vivo monitoring device outside the shielded counting room 

(Watts et al. 1995). It has three germanium detectors facing downward and two NaI detectors on the 

sides of the bed. As described below, it can also be used as a backup chest counter, if necessary, 

due to its 3-to-50% efficient germanium detectors (Watts et al. 1995). The detectors are in a solid 

center shield. The shield is designed so there is a minimum of 4 in. of lead around each detector. 

During this time, the 5.25- by 0.0625-in. NaI - 2-in. CsI phoswich detector was used to conduct chest 

counts. Some time between 1977 and 1994, another phoswich detector with similar dimensions was 

procured and used until three 70% efficient germanium detectors were installed in the counting room 

in 1994 (Hembree 2004; McLaughlin 2004). A chair counting geometry was reinitiated at that time, 

with the individual positioned with their back reclined 45 degrees from the vertical. The three 

germanium detectors are positioned with two detectors over the right lung (one over the upper portion 

and the other over the lower portion) and one over the left lung. The third detector is over the upper 

portion of the left lung. 

In addition to the above, there are three other detector arrangements in the WBC facility. A 

germanium thyroid counter can be set up using one of the detectors used for lung counting. The 

individual is in the same chair used for lung counting, but the chair is inclined at a 25-degree angle. 

Other organ-specific geometries can be used as necessary (skull, liver, hand, etc.). The 5.25- by 

0.0625-in. NaI - 2-in. CsI phoswich and the SBC (in a fixed-bed position) are backup systems that 

could be used for lung counts if needed. 

Until 1994, field health physicists selected individuals for in vivo monitoring. At that time, the internal 

dosimetry group became responsible for the selection of personnel for in vivo monitoring with 

guidance from field health physicists. Berger (2003) indicated that the in vivo monitoring program 

may not have been effective for monitoring ORNL personnel until the late 1980s because of reliance 

on field health physicists to determine the individuals to be monitored. 

5.3.3

Many ORNL workers had measured body burdens from intakes of 137Cs from nonoccupational 

sources (e.g., fallout and consumption of local venison). At present, ORNL uses a value of 20 nCi 

137Cs in a whole-body count for consumers of venison as the decision level to follow up or conduct a 

Cesium Counting Artifact
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 26 of 70

dose assessment (ORNL 2002). However, the Laboratory has never used it to negate an individual’s 

occupational dose that was received on the site (McLaughlin 2004). 

Fallout affects everyone in North America, and body burdens of 137Cs measurable in whole-body 

counters were common in the 1960s and 1970s. NCRP Report 94 (NCRP 1988) provides mean body 

burdens of 137Cs for the United States for the years most likely to produce interference with 

occupational whole-body count results. Table 5-16 lists these values. 

Table 5-16. Mean body burdens of 137Cs from fallout in the 

United States. 

Year Body burden (nCi) Year Body burden (nCi)

1953 0.27 1966 9.7

1954 1.1 1967 5.6

1955 2.2 1968 3.5

1956 4.3 1969 2.7

1957 5.1 1970 2.7

1958 6.5 1971 2.7

1959 8.1 1972 2.7

1960 6.8 1973 2.7

1961 4.6 1974 1.6

1962 6 1975 1.1

1963 11 1976 1.6

1964 19 1977 1.1

1965 16

5.3.4

As stated above, the thin (5-in.-diameter by 0.0625 in.) NaI (Tl) crystal array used in 1963 to detect 

low-energy photons emitted primarily from transuranics could detect approximately 40 nCi of 238Pu, if 

there had been a preexposure chest count. If there was no preexposure count, the detection 

capability was approximately 80 nCi (Morgan et al. 1963).

Minimum Detectable Activity

Table 5-17 lists MDAs for the HPGe array of six individual detectors in the aluminum block (used 

around early- to mid-1980s).

Table 5-17. MDA values for six-detector 

HPGe array (Berger and Lane 1981).

Nuclide Organ MDA (nCi) 239Pu Lung 21.5 241Am Lung 1.1 125I Thyroid 0.04 137Cs Lung 16.4 153Gd Lung 0.87 152Eu Lung 0.71 233U Lung 1.42 57Co Lung 0.94

Tables 5-18, 5-19, and 5-20 list MDA values for the ORNL SBC, germanium thyroid counter, and fixed 

bed counter, respectively (Watts et al. 1995). MDA values are calculated for each individual 

measurement made using the germanium lung counter, SBC, and fixed bed counter. Figure 5-1 is a 

plot of MDA values for different nuclides for the ORNL germanium lung counter as a function of chest 

wall thickness. Figure 5-2 shows MDA values for 238Pu and 241Am for the ORNL phoswich lung 

counter as a function of chest wall thickness. Table 5-21 lists the frequencies of in vivo counting since 

1960 and, for the various detectors, a summary of the best estimates of the applicable periods of use.
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 27 of 70

Table 5-18. MDA values for ORNL scanning bed counter (NaI/Ge) (Watts 1995).

Isotope MDA (nCi) Isotope MDA (nCi) Isotope MDA (nCi)

51Cr 24/30 54Mn 2.1/2.4 57Co 4.9/5.9

59Fe 3.9/4.5 60Co 2.2/1.9 65Zn 4.3/6.0

88Rb 12/15 88Y 1.8/1.9 95Nb 2.1/2.7

95Zr 3.7/4.1 99mTc 4.4/5.3 103Ru 2.6/2.9

106Ru 21/30 109Cd NG/224* 113Sn 3.4/4.0

124Sb 2.2/3.1 131I 2.7/3.1 133Ba 3.7/4.5

133I 2.7/2.9 134Cs 2.2/3.0 137Cs 2.4/2.8

139Ce 4.3/4.6 140Ba 9.2/9.1 140La 2.3/2.4

141Ce 7.7/8.4 144Ce 36/43 152Eu 8.5/9.7

153Gd 21/21 154Eu 6.1/7.2 212Pb 6.3/7.6

214Bi 5.8/9.8 235U 5.9/6.1

*NG = not given

Table 5-19. Current MDA 

values for ORNL germanium 

thyroid counter.

Isotope MDA (nCi)

125I 0.06

131I 0.12

Table 5-20. Current MDA 

values for ORNL fixed bed 

counter.

Isotope MDA (nCi)

60Co 0.31

144Ce 6.28

Figure 5-1. Current MDA values (nCi) for ORNL germanium 

lung counter versus chest-wall thickness (cm). (Watts et al. 

1995) 

241Am 238Pu

Figure 5-2. Current MDA values (nCi) for ORNL phoswich lung counter versus chest-wall thickness 

(cm) for 241Am and 238Pu (Watts et al. 1995). 
Effective Date: 05/30/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0012-5 Page 28 of 70

Table 5-21. Frequency of in vivo monitoring. Monitoring type Applicable period Frequency

Lung counting

1960 – 1961

4- x 4-in. NaI (Tl)

No set frequency. (Period of facility development –limited routine human 

counting, typically of incidents)

1961 – 1962

8- x 4-in. NaI (Tl)

No set frequency. (Development activities to change over to larger 

detector – typically incidents)

1962 – 1967-71

8- x 4-in. NaI (Tl)

No set frequency. (Period spent obtaining baseline counts onsite 

employees and incident counts)

1965 - ??

8- x 4-in. NaI (Tl)

Frequency stated in Annual Report (see Table 5-12), but discussions with 

site personnel indicate that it was not consistently implemented.

ca 1967-71 to 1976

8- x 4-in. NaI (Tl) and 9- x 9-in. NaI 

(Tl)

Large detector was added below counting table and used when 

conducting stationary lung counts only.

1962 – 1976

5-in. dia. x 1/16-in. NaI (Tl)

No set frequency. (Thin NaI crystals used to quantify low energy photons.)

1976 – 1994

5.25- x 1/16-in. NaI - 2- CsI 

phoswich

Monitoring frequency was not effectively implemented. This phoswich 

detector was used to monitor for low energy photons.

1980 – 1987

Germanium array and phoswich

Monitoring frequency was not effectively implemented. Detectors used 

together.

1992 – present

combination germanium and NaI

Fixed Bed Counter is SBC in fixed position. Used only to measure fission 

and activation products (e.g., 60Co and 144Ce) immediately after 

assimilation.

1994 – present

Germanium system

Varies from annual to semiannual depending on exposure potential set by 

Internal Dosimetrist.

1994 – present

5.25- x 1/16-in. NaI - 2- CsI 

phoswich

Phoswich detector can be used, but germanium detection system noted 

above is routinely used.

Whole-body counting

1960 – 1961

4- x 4-in. NaI (Tl)

No set frequency. (Period of facility development –limited routine human 

counting, typically incidents)

1961 – 1962

8- x 4-in. NaI (Tl)

No set frequency (Development activities to change to larger detector –

typically incidents)

1963 – 1976

8- x 4-in. NaI (Tl)

Table 5-12 Selection criteria used in 1965 for WBC

1992 – present

combination germanium and NaI

SBC. Monitoring frequency implemented in 1994 to include annual and 

semiannual counts depending on exposure potential set by Internal 

Dosimetrist.

Thyroid counter (and 

other organs)

1963 - ??

3- x 3-in. NaI (Tl)

Incident counting and research. May have been used for frequent 

monitoring of workers potentially exposed to iodine, but this information 

was not located.

1994 – present 

germanium (uses one of three 

germanium detectors used for lung 

counting)

Typically, site research personnel would be placed on this program and 

monitored on semiannual frequency, as well as persons involved in known 

or potential iodine intake. 

5.4 USING AIR MONITORING DATA TO ASSIGN BEST ESTIMATES OF DOSE

Information about internal dose at Oak Ridge National Laboratory (ORNL) from 1943 through 1947 is 

somewhat limited, with no bioassay results appearing in the DOE worker files. However, there are 

fairly extensive air-monitoring records from 1944 to 1947 that are suitable for assigning best estimates 

of dose. The general protocol presented in Internal Dose Overestimates for Facilities with Air 

Sampling Programs (ORAUT 2005) is used with the following modifications:

• The approach applies only to best estimates of doses received in the Februay 1, 1943, to 

December 31, 1947, period. Construction of the pilot plant began on February 1, 1943, and it 

was placed into operation on November 4, 1943 (Thompson 1963). Although air-monitoring 

data from before June 1944 was not available, it is assumed that airborne rates would have 

generally increased from the start of operations until the time when routine airborne activity 

levels are available.

• For the alpha-emitter category, the radionuclides are limited to plutonium and uranium.
Effective Date: 05/30/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0012-5 Page 29 of 70

• A constant distribution is used in IREP for the doses rather than a lognormal distribution. 

Attachment 5B discusses the technical justification for this approach.
Effective Date: 05/30/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0012-5 Page 30 of 70

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for 1974, ORNL-5055, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Auxier, J. A., 2003. Personal communications with Mike Bollenbacher, Knoxville, Tennessee, October 

15.

Berger, C. D., and R. E. Goans, 1981, “A Comparison of the NaI-CsI Phoswich and a Hyperpure 

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40, pp. 535-542.

Berger, C. D., and B. H. Lane, 1981, Performance of a Large Hyperpure Germanium Detector Array 

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Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 33 of 70

ATTACHMENT 5A

CALCULATION OF MDAS FROM HISTORICAL DATA

TABLE OF CONTENTS

Section

5A.1 Recovery of Historical Data.......................................................................................................

Page

34

5A.1.1 Data Description...............................................................................................................34

5A.1.2 Data Classification............................................................................................................ 34

5A.1.3 Synopsis of Recovered Data ............................................................................................ 35

5A.2 Calculation Method for Sample MDAs.......................................................................................35

5A.2.1 Standard Deviation of Background Counts....................................................................... 36

5A.2.2 Background Count Time................................................................................................... 37

5A.2.3 Instrument Efficiency ........................................................................................................ 37

5A.2.4 Yield................................................................................................................................. 37

5A.2.5 Volume of Aliquot Analyzed.............................................................................................. 37

5A.2.6 Volume of 24-Hour Sample .............................................................................................. 37

5A.3 Calculation Method for MDAs for Various Periods Between 1945 and 1988 ............................. 37

5A.4 Additional Considerations .........................................................................................................38

LIST OF TABLES

Table

5A-1 Data used to determine historical MDAs for radionuclides in urine and feces.........................

Page

36

5A-2 MDAs (dpm/24-hr sample) calculated from recovered data for radionuclides......................... 39

LIST OF FIGURES

Figure

5A-1 MDA data for americium-241 in urine.....................................................................................

Page

40

5A-2 MDA data for curium-244 in urine...........................................................................................41

5A-3 MDA data for Cs-137 in urine................................................................................................. 43

5A-4 MDA data for gross alpha in urine..........................................................................................44

5A-5 MDA data for gross beta in urine............................................................................................46

5A-6 MDA data for H-3 in urine.......................................................................................................47

5A-7 MDA data for Pm-147 in urine................................................................................................49

5A-8 MDA data for Pu-238 in urine................................................................................................. 50

5A-9 MDA data for Pu-239 in urine................................................................................................. 52

5A-10 MDA data for Pu-241 in urine................................................................................................. 53

5A-11 MDA data for rare earths in urine ...........................................................................................55

5A-12 MDA data for Sr-90 in urine....................................................................................................56

5A-13 MDA data for U-233 in urine...................................................................................................58

5A-14 MDA data for U-238 in urine...................................................................................................59

5A-15 MDA data for Am-241 in feces ...............................................................................................61

5A-16 MDA data for gross alpha in feces .........................................................................................62

5A-17 MDA data for Pu-239 in feces ................................................................................................64

5A-18 MDA data for Th-232 in feces ................................................................................................65
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 34 of 70

5A.1 RECOVERY OF HISTORICAL DATA

A large amount of raw bioassay data exists at the ORNL complex. A portion of these data was 

transcribed into an electronic database hosted on ORNL equipment and maintained by ORNL 

personnel.

As part of this dose reconstruction effort, this database was queried for historical urinalysis records on 

October 30, 2003. A supplemental query was performed on December 12, 2003. These queries 

requested information on the samples and the instrumentation used in the analysis. No personal 

information or individual dose results were requested or received.

These database queries generated 66,204 records from the period between 1945 and 1988. These 

records were compiled into a single database. A data validation and comment field was added to 

each record. Then the records were examined and classified according to usability.

5A.1.1

The database queries retrieved 17 parameters from the ORNL internal dosimetry database:

Data Description

• NUCLIDE - A unique abbreviated identifier for specific nuclides

• SAMPLEID - A unique identifier for individual samples

• SAMPLETYPE – A one-character identifier designating the type of sample (urine, fecal, etc.)

• ALIQUOTVOL – The volume or mass of the sample analyzed

• SAMPLEVOL – The sample volume or mass collected

• COUNTDATE – The date the aliquot was counted

• STARTEDATE – The date sample processing began

• STOPDATE – The date sample processing ended

• RECOVERY – The chemical yield, in percent

• EFFICIENCY – The detector’s counting efficiency, in percent

• EFF_FACTOR – The reciprocal of the detector’s counting efficiency

• COUNTMIN – The sample count time, in minutes

• T_BKG_RATE – The total background count rate, in counts per minute

• BKG_MIN – The background count time, in minutes

• BKG_RGNT_R – The count rate of any reagent used

• BKG_COUNTS – The number of background counts

• MDA – The minimum detectable activity

5A.1.2

The 66,204 individual records in the database were classified into one of four categories:

Data Classification

• Duplicate Records - 272 duplicate records were found. These were marked with a “D” in the 

record’s validation field.

• Usable without Modification - 58,972 of the records collected (89%) were usable without 

qualification. 

• Usable with Modification - Some of the records collected contained input errors that were 

easily identified. Examples include a nuclide entry of C-137, instrument efficiencies listed in 

fractions instead of percent, and dates from the turn of the century. Other information in the 
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 35 of 70

database enabled correction of some of these records. Such modifications were designated 

by placing an “M” in the records validation field and entering the reason for the change in the 

comment field. These records comprise 3% (2,193) of the database.

• Unusable Records - 4,767 of the records recovered (7%) were determined to be unusable in 

their present form. Most of these records are incomplete or contain gross errors. These were 

marked with an R in the validation field. An explanation for a record’s rejection was entered in

the comment field.

The resulting database provided all the raw data used to determine the MDAs of radionuclides in 

urinalysis calculated in this attachment.

5A.1.3

The database queries recovered 61,165 usable records. Of these, 59,288 are clearly records of 

urinalyses and 1,859 are records of fecal analyses. Special analyses made up the remaining 18 

records. Seventy-five different analyte labels are listed, although some are clearly different names for 

the same analyte (e.g. GA, GAO, and G_ALPHA are all gross alpha measurements). Table 5A-1 lists 

the recorded analyses performed by radionuclide. 

Synopsis of Recovered Data

5A.2 CALCULATION METHOD FOR SAMPLE MDAS

Equations 5A-1 and 5A-2 show the derivation of the estimated MDA, which was calculated from 

61,165 analytical records. MDAs were calculated using the method in Brodsky (1992):





 



 ×

×

×+ =

dis

24cnts h sample T (min) k

3 SD4.65 (cnts) MDA 

b

b (Eq. 5A-1)

where

MDA = Minimum Detectable Activity (dpm/24 h sample)

SDb = Standard deviation of the total background counts (cnts/min)

Tb = Duration of the background count (min)

k = Calibration factor (cnts•24 h sample/dis)







 





 



 



 



 



 ×

24 h sample

mL V

 V (mL) x Yield x 

dis

cnts = InstrEff

dis

24cnts h sample k

r

a (Eq. 5A-2)

where:

InstrEff = Instrument efficiency (cnts/dis)

Yieldi = Fractional chemical recovery

Va = Volume of the aliquot analyzed (ml for urine, g for feces) 

Vr = Volume of the sample submitted, or the volume of a standard 24-hr sample from 

reference man (1,400 ml/24-hr sample for urine, 135 g/24-hr sample for feces) 

(ICRP Publication 23 [ICRP Year] default values for 24-hr voids), whichever is 

larger

To use this equation one must know, or be able to calculate, the standard deviation of background 

counts, the duration of the count, the yield of any chemical extractions used, the efficiency of the 
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 36 of 70

Table 5A-1. Data used to determine historical MDAs for radionuclides in urine and feces. 

Nuclide Fecal Urinalysis Row totals

Totals 1,859 59,288 61,147

Alpha 1 3 4

Am-241 130 5,540 5,670

Am-243 6 6 12

As-74 7 7

BG 2 2

Bk-249 2 12 14

Br-82 2 2

Br-83 2 2

C-14 11 11

Ca-45 4 4

Ce-144 4 33 37

Cf-249 2 1 3

Cf-252 3 11 14

Cl-36 1 1

Cm-242 1 11 12

Cm-244 26 273 299

Co-60 2 81 83

Cs-134 1 1

Cs-137 4 3,557 3,561

Fe-59 1 8 9

GA 69 1,875 1,944

G-alpha 6 6

GAO 154 2,771 2,925

GB 278 278

GBO 46 46

H-3 2,070 2,070

I-131 41 41

Mn-54 2 2

Mo-99 1 1

Na-24 3 3

Nb-95 3 3

Np-237 1 54 55

P-32 166 166

Pa-231 1 54 55

Pa-233 3 13 16

Pa-234 1 1

Nuclide Fecal Urinalysis Row totals

Pb-210 2 2

Pm-147 16 64 80

Po-210 9 57 66

Pu-238 4 61 65

Pu-239 124 15,352 15,476

Pu-241 1 111 112

Pu-242 41 41

Ra-226 1 332 333

Ra-228 1 1

RE 39 698 737

Ru-103 1 1

Ru-106 65 65

S-35 10 10

Sb-125 1 1

Sm-151 11 11

Sr-85 1 1

Sr-89 37 37

Sr-90 142 12,751 12,893

T 1 1

Tc-99 19 19

Th-232 1,079 46 1,125

Tl-201 1 1

Tl-204 1 1

Tm-170 1 5 6

TPU 1 1

TRE 2 197 199

TRE B 1 1

TRE-B 4 4

U-232 1 1

U-233 3 826 829

U-235 3 3

U-238 25 11,409 11,434

U-239 11 11

Y+RE 157 157

Y-88 5 5

Y-90 31 31

Zn-65 7 7

Zr-95 1 19 20

detector used, and the amount of the sample analyzed. The method of recovering this information 

from the historical database is discussed in the following sections.

5A.2.1

The data field BKG_COUNTS contained recorded background counts for a sample and was the 

preferred source of background count information. When a record’s BKG_COUNTS field was blank, 

the background count information was calculated as the product of the T_BKG_RATE and BKG_MIN 

fields. If the BKG_MIN field was blank, but the T_BKG_RATE field contained a value, the background 

count information was calculated as the product of the T_BKG_RATE and COUNTMIN fields.

Standard Deviation of Background Counts
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 37 of 70

The variability in repeat counts, such as background counts, follows the Poisson distribution (Evans 

1955, Chapter 26). It is a property of the Poisson distribution that its standard deviation is the square 

root of its mean value. Therefore, in this analysis, the standard deviation of the total background 

counts was calculated as the square root of the total background counts.

5A.2.2

The data field BKG_MIN contained the recorded background count time for a sample and was the 

preferred source of background count time information. If the BKG_MIN field was blank, the time in 

the COUNTMIN field was used. This last method of estimating the background count time is likely to 

be claimant-favorable because in almost all cases where there is paired data, the COUNTMIN value 

is less than the BKG_MIN value. This produces a higher MDA when substituted into Equation 1.

Background Count Time

5A.2.3

Information on instrument efficiencies was contained in two fields. The EFFICIENCY field contained 

the percent of disintegrations detected by the instrument. The EFF_FACTOR field contained the 

reciprocal of the EFFICIENCY field.

Instrument Efficiency 

5A.2.4

The chemical yield for a given radiochemical analysis was contained in the RECOVERY data field. 

This value was given in percent recovery. 

Yield

Many of the records had blank recovery fields. However, due to the repetitive nature of the chemical 

extractions used, application of an average, radionuclide-specific, chemical recovery to records with 

no recovery value was judged to be acceptable. This approach is supported further by the lack of 

variability observed in daily runs on known standards recorded in laboratory logbooks recovered from 

the late-1950s through the mid-1970s.

5A.2.5

The ALIQUOTVOL field usually contained a value for the amount of material subjected to 

radioanalysis. When this field was blank, it was assumed that the entire sample provided was used in 

the analysis. The amount of the sample was recorded in the SAMPLEVOL field.

Volume of Aliquot Analyzed

5A.2.6

Excretion rates and concentrations vary greatly within a 24-hr period and between individuals. The 

models developed for calculating intakes from in vitro analyses are based on 24-hr samples. The 

daily urine excretion rate is 1,400 ml/24-hr sample; for feces, this value is 135 g/24-hr sample.

Volume of 24-Hour Sample

Often, the volume analyzed (Va) was less than the reference excretion rate from standard man. When 

this was observed, the calculated activity in the sample was normalized to the reference man volume 

to compensate. This adjustment increased the MDA and, therefore, would be claimant-favorable.

5A.3 CALCULATION METHOD FOR MDAS FOR VARIOUS PERIODS BETWEEN 1945 AND 

1988

Once the MDAs were calculated for individual samples, they were segregated by radionuclide and 

imported into Excel spreadsheets. Using Excel, the values were plotted for visual inspection. Figures 
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 38 of 70

5A-1 through 5A-18 at the end of this attachment show plots of data the was obtained and calculated 

for the following radioisotopes: (urine) 241Am, 244Cu, 137Cs, gross alpha, gross beta, 3

H, 147Pm, 238Pu, 

239Pu, 241Pu, rare earths, 90Sr, 233U, 238U, and (feces) 241Am, gross alpha, 239Pu, 232Th. In the first plot, 

the MDA values were plotted against the date of their analysis. This provided a visual representation 

of individual MDAs over time. The density of the available data and a gross approximation of typical 

sample MDAs for a given period were quickly observed using this format.

Two bar charts were created for each radionuclide. One shows the annual average MDAs between 

1945 and 1988. The other shows the number of analytical results recovered for each year in the 

same period. These two charts provide additional information on gross trends exhibited by sample 

MDAs, and on the temporal distribution of the data.

These visual aids were used to establish general periods during which sample MDAs were similar. 

The sample MDAs in these periods were grouped as similar populations. Using Excel’s statistical 

functions, the following statistical parameters for each group of samples were determined:

• The population size (N)

• The 50th percentile value of the population,

• The population’s arithmetic mean

• The arithmetic standard deviation of the population

• The 95% confidence level on the mean

The 95% confidence levels on the mean for each radionuclide are the recommended MDAs for that 

radionuclide during the period evaluated. 

5A.4 ADDITIONAL CONSIDERATIONS 

The recovered data indicate the analytical MDAs tended to remain fairly consistent for years. Abrupt 

changes in the MDAs were identified for groups of radionuclides during specific years. Following the 

change, the MDAs remained generally consistent in the several succeeding years. This “step-wise” 

pattern allowed MDAs from several years to be grouped as one MDA for a specified period. Table 5-9 

highlights these groupings by enclosing similar MDAs in a box.

Some MDAs changed frequently, creating a pattern of similar MDAs in short adjacent periods. When 

this occurred, the largest estimated MDA from that period was selected to be the representative MDA 

for the entire period. The tritium (3

H) MDA listed between 1968 and 1981 in Table 5-9 is an example 

of this adjustment.

Most recovered data contained isotope-specific entries in the NUCLIDE field. However, the ability to 

differentiate routinely between isotopes of the same radioelement did not always exist before 1989. 

Many of the isotopic assignments were based on process knowledge. Therefore, during the period 

prior to 1989, the MDA for some isotopes are reported as the MDA for the radioelement instead of the 

radionuclide. For example, the alpha emitters 238Pu and 239

Pu are reported separately in the 

database. They have been combined in this TBD as plutonium. NOTE: 241Pu is reported separately 

because it is a beta emitter that can be measured separately from the alpha emitting isotopes of 

plutonium. If the recovered data did not extend to 1989, the MDA calculated from the last known data 

was extrapolated forward (see Table 5-9). 
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 39 of 70

Table 5A-2. MDAs (dpm/24-hr sample) calculated from recovered data for radionuclides. 

Year

Urine MDAs (dpm/24h sample) Fecal MDAs (dpm/24h sample)

Gross 

alpha

Gross 

beta

Am241Cm244Cs137 H-3I131Np237Pm147 Plutonium PoloniumPu241Ra226Rare 

earths

Ru106Sr-89 + 

Sr-90 Uranium

Gross

alpha

Am241Cm244 PlutoniumTh232

1943

1944

1945

1946

1947 0.33

1948 0.40 0.965

1949 0.34 0.998

1950 24.34 0.19 53.95 1.401 0.176

1951 29.54 0.18 33.34 0.036 22.34 3.600

1952 74.44 0.27 34.93 0.23 25.44 31.41 27.50 5.988

1953 0.17 29.49 0.19 2.98 30.45 27.01 6.196

1954 0.25 351.14 32.90 0.17 9.42 0.32 41.97 29.00 6.296

1955 0.22 960.93 26.52 0.19 33.32 0.87 0.142 26.76 6.810

1956 0.23 889.95 0.25 32.05 0.22 57.11 0.16 0.238 32.75 7.651

1957 0.69 998.73 0.44 30.61 1.07 0.39 0.80 0.272 28.08 5.257 0.055 0.072 0.226

1958 0.35 1385.09 0.42 38.61 0.26 0.19 0.34 0.255 47.99 7.423 0.728 0.537

1959 0.31 1321.55 0.35 312.8 0.10 587 0.23 0.29 331.74 0.359 36.83 5.636 0.647 0.303

1960 0.19 938.01 0.11 147.3 0.27 196 0.18 10.67 0.17 167.87 61.85 19.60 5.740 0.568 0.573

1961 0.21 290.98 42.62 71382 0.18 12.74 0.29 50.04 5.86 5.686 0.512 0.424

1962 0.32 252.50 641672 0.14 54.69 8.88 8.687 0.422

1963 0.45 146.59 72.18 393539 0.49 25.81 0.10 87.30 14.79 7.747 1.075 0.789 0.633

1964 0.24 198.28 0.57 62.09 214873 0.19 78.91 12.26 2.561 0.118 1.052

1965 0.46 140.95 0.21 0.17 55.66 124229 353.2 0.08 0.14 63.41 13.78 0.992 0.848 0.208 0.172 0.163 0.112

1966 0.12 0.16 52.71 153077 0.08 6.58 0.128 0.447 0.268

1967 0.082 0.099 75.55 150699 35718 0.08 7.30 0.405 0.769

1968 0.056 36.17 0.073 0.080 55.59 54872 1131 0.07 0.07 2.55 25.76 15.70 49.40 6.68 0.112 0.079 1.059 0.349

1969 0.066 38.82 0.099 0.082 50.80 53036 1281 0.09 23.10 0.07 31.75 7.66 56.96 46.95 4.73 0.094 0.768 0.306 0.536

1970 0.038 32.31 0.074 0.096 44.32 48763 1140 0.05 0.05 5.67 4.05 0.071 0.097 0.271 0.097

1971 37.06 0.072 49037 0.06 677 0.06 6.00 2.82 4.33 0.076 0.230 0.351

1972 0.18 0.046 75.05 45882 0.04 24.06 0.05 9.18 5.00 3.95 0.061 0.093 0.035 0.033

1973 0.10 0.039 50535 6.44 0.04 6.94 5.44 4.14 0.071 0.105 0.060

1974 0.093 46154 5.72 0.04 4.53 4.86 4.80 0.054 0.543 1.073 1.040

1975 0.040 0.032 44192 0.04 0.46 1.49 3.26 0.054 0.047 0.129 0.073 0.046

1976 0.050 0.080 46.12 45554 0.04 1.84 0.014 2.24 3.58 0.057 0.173 0.053

1977 0.037 36984 2.82 0.04 9.65 0.049 3.62 0.053 0.121 0.100

1978 0.032 0.016 31838 0.14 0.03 8.16 3.39 3.30

1979 0.030 0.046 35406 0.03 0.03 0.23 12.40 3.06 0.029 0.032

1980 0.032 0.036 48449 3.05

1981 0.015

1982

1983

1984 0.026

1985

1986 0.07

1987 0.07

1988 0.028
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 40 of 70

Visual Comparison of Data to Estimated Am-241 MDAs in Urine

1 E-3

1 E-2

1 E-1

1 E+0

1 E+1

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA (95%CL on Mean)

50th Percentile

Figure 5A-1. MDA data for americium-241 in urine (part 1 of 3).

Annual Average MDAs for Am-241 in Urine 

Calculated from Recovered Data Between 1945 and 1990

0

0.05

0.1

0.15

0.2

0.25

0.3

0.35

0.4

0.45

0.5

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-1. MDA data for americium-241 in urine (part 2 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 41 of 70

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Am-241 in Urine 

0

100

200

300

400

500

600

700

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-1. MDA data for americium-241 in urine (part 3 of 3).

Visual Comparison of Data to Estimated Cm-244 MDAs in Urine

1 E-3

1 E-2

1 E-1

1 E+0

1 E+1

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA of 0.2 (95%CL on Mean)

50th Percentile of 0.1

Figure 5A-2. MDA data for curium-244 in urine (part 1 of 3) (some later tables are “in 

feces”).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 42 of 70

Annual Average MDAs for Cm-244 in Urine 

Calculated from Recovered Data Between 1945 and 1990

0

0.1

0.2

0.3

0.4

0.5

0.6

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-2. MDA data for curium-244 in urine (part 2 of 3).

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Cm-244 in Urine 

0

10

20

30

40

50

60

70

80

90

100

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-2. MDA data for curium-244 in urine (part 3 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 43 of 70

Visual Comparison of Data to Estimated Cs-137 MDAs in Urine

1 E+0

1 E+1

1 E+2

1 E+3

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

MDA (dpm/24 h)

Estimated MDA For the Period

50th Percentile

Figure 5A-3. MDA data for Cs-137 in urine (part 1 of 3).

Annual Average MDAs for Cs-137 in Urine 

Calculated from Recovered Data Between 1945 and 1990

0

50

100

150

200

250

300

350

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-3. MDA data for Cs-137 in urine (part 2 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 44 of 70

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Cs-137 in Urine 

0

100

200

300

400

500

600

700

800

900

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-3. MDA data for Cs-137 in urine (part 3 of 3).

Visual Comparison of Data to Estimated Gross Alpha MDAs in Urine

1 E-2

1 E-1

1 E+0

1 E+1

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA (95%CL on Mean)

50th Percentile

Figure 5A-4. MDA data for gross alpha in urine (part 1 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 45 of 70

Annual Average MDAs for Gross Alpha in Urine 

Calculated from Recovered Data Between 1945 and 1990

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-4. MDA data for gross alpha in urine (part 2 of 3).

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Gross Alpha in Urine 

0

500

1000

1500

2000

2500

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-4. MDA data for gross alpha in urine (part 3 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 46 of 70

Visual Comparison of Data to Estimated Gross Beta MDAs in Urine

1 E+1

1 E+2

1 E+3

1 E+4

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA (95%CL on Mean)

50th Percentile

Figure 5A-5. MDA data for gross beta in urine (part 1 of 3).

Annual Average MDAs for Gross Beta in Urine 

Calculated from Recovered Data Between 1945 and 1990

0

200

400

600

800

1000

1200

1400

1600

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-5. MDA data for gross beta in urine (part 2 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 47 of 70

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Gross Beta in Urine 

0

10

20

30

40

50

60

70

80

90

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-5. MDA data for gross beta in urine (part 3 of 3).

Visual Comparison of Data to Estimated H-3 MDAs in Urine

1 E+3

1 E+4

1 E+5

1 E+6

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA (95%CL on Mean)

50th Percentile

Figure 5A-6. MDA data for H-3 in urine (part 1 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 48 of 70

Annual Average MDAs for H-3 in Urine 

Calculated from Recovered Data Between 1945 and 1990

0

100000

200000

300000

400000

500000

600000

700000

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-6. MDA data for H-3 in urine (part 2 of 3).

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for H-3 in Urine 

0

50

100

150

200

250

300

350

400

450

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-6. MDA data for H-3 in urine (part 3 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 49 of 70

Visual Comparison of Data to Estimated Pm-147 MDAs in Urine

1 E+0

1 E+1

1 E+2

1 E+3

1 E+4

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA of 396 (95%CL on Mean)

50th Percentile of 40

Figure 5A-7. MDA data for Pm-147 in urine (part 1 of 3).

Annual Average MDAs for Pm-147 in Urine 

Calculated from Recovered Data Between 1945 and 1990

0

100

200

300

400

500

600

700

800

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-7. MDA data for Pm-147 in urine (part 2 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 50 of 70

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Pm-147 in Urine 

0

2

4

6

8

10

12

14

16

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-7. MDA data for Pm-147 in urine (part 3 of 3).

Visual Comparison of Data to Estimated Pu-238 MDAs in Urine

1 E-3

1 E-2

1 E-1

1 E+0

1 E+1

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA (95%CL on Mean)

50th Percentile

Figure 5A-8. MDA data for Pu-238 in urine (part 1 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 51 of 70

Annual Average MDAs for Pu-238 in Urine 

Calculated from Recovered Data Between 1945 and 1990

0

0.1

0.2

0.3

0.4

0.5

0.6

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-8. MDA data for Pu-238 in urine (part 2 of 3).

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Pu-238 in Urine 

0

1

2

3

4

5

6

7

8

9

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-8. MDA data for Pu-238 in urine (part 3 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 52 of 70

Visual Comparison of Data to Estimated Pu-239 MDAs in Urine

1 E-3

1 E-2

1 E-1

1 E+0

1 E+1

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA (95%CL on Mean)

50th Percentile

Figure 5A-9. MDA data for Pu-239 in urine (part 1 of 3).

Annual Average MDAs for Pu-239 in Urine 

Calculated from Recovered Data Between 1945 and 1990

0

0.05

0.1

0.15

0.2

0.25

0.3

0.35

0.4

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-9. MDA data for Pu-239 in urine (part 2 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 53 of 70

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Pu-239 in Urine 

0

200

400

600

800

1000

1200

1400

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-9. MDA data for Pu-239 in urine (part 3 of 3).

Visual Comparison of Data to Estimated Pu-241 MDAs in Urine

1 E-1

1 E+0

1 E+1

1 E+2

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA of 9.2 (95%CL on Mean)

50th Percentile of 6.7

Figure 5A-10. MDA data for Pu-241 in urine (part 1 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 54 of 70

Annual Average MDAs for Pu-241 in Urine 

Calculated from Recovered Data Between 1945 and 1990

0

5

10

15

20

25

30

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-10. MDA data for Pu-241 in urine (part 2 of 3).

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Pu-241 in Urine 

0

5

10

15

20

25

30

35

40

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-10. MDA data for Pu-241 in urine (part 3 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 55 of 70

Visual Comparison of Data to Estimated Rare Earth MDAs in Urine

1 E-1

1 E+0

1 E+1

1 E+2

1 E+3

1 E+4

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA (95%CL on Mean)

50th Percentile

Figure 5A-11. MDA data for rare earths in urine (part 1 of 3).

Annual Average MDAs for Rare Earths in Urine 

Calculated from Recovered Data Between 1945 and 1990

0

50

100

150

200

250

300

350

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-11. MDA data for rare earths in urine (part 2 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 56 of 70

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Rare Earths in Urine 

0

100

200

300

400

500

600

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-11. MDA data for rare earths in urine (part 3 of 3).

Visual Comparison of Data to Estimated Sr-90 MDAs in Urine

1 E-1

1 E+0

1 E+1

1 E+2

1 E+3

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA (95%CL on Mean)

50th Percentile

Figure 5A-12. MDA data for Sr-90 in urine (part 1 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 57 of 70

Annual Average MDAs for Sr-90 in Urine 

Calculated from Recovered Data Between 1945 and 1990

0

10

20

30

40

50

60

70

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-12. MDA data for Sr-90 in urine (part 2 of 3).

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Sr-90 in Urine 

0

200

400

600

800

1000

1200

1400

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-12. MDA data for Sr-90 in urine (part 3 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 58 of 70

Visual Comparison of Data to Estimated U-233 MDAs in Urine

1 E-2

1 E-1

1 E+0

1 E+1

1 E+2

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA (95%CL on Mean)

50th Percentile

Figure 5A-13. MDA data for U-233 in urine (part 1 of 3).

Annual Average MDAs for U-233 in Urine 

Calculated from Recovered Data Between 1945 and 1990

0

1

2

3

4

5

6

7

8

9

10

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-13. MDA data for U-233 in urine (part 2 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 59 of 70

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for U-233 in Urine 

0

50

100

150

200

250

300

350

400

450

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-13. MDA data for U-233 in urine (part 3 of 3).

Visual Comparison of Data to Estimated U-238 MDAs in Urine

1 E-3

1 E-2

1 E-1

1 E+0

1 E+1

1 E+2

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Estimated MDA For the Period

50th Percentile

Figure 5A-14. MDA data for U-238 in urine (part 1 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 60 of 70

Annual Average MDAs for U-238 in Urine 

Calculated from Recovered Data Between 1945 and 1990

0

1

2

3

4

5

6

7

8

9

10

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-14. MDA data for U-238 in urine (part 2 of 3).

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for U-238 in Urine 

0

100

200

300

400

500

600

700

800

900

1000

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-14. MDA data for U-238 in urine (part 3 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 61 of 70

Visual Comparison of Data to Estimated Am-241 MDAs in Feces

1 E-2

1 E-1

1 E+0

1 E+1

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA of 0.5 (95%CL on Mean)

50th Percentile of 0.2

Figure 5A-15. MDA data for Am-241 in feces (part 1 of 3).

Annual Average MDAs for Am-241 in Feces 

Calculated from Recovered Data Between 1945 and 1990

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-15. MDA data for Am-241 in feces (part 2 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 62 of 70

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Am-241 in Feces 

0

5

10

15

20

25

30

35

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-15. MDA data for Am-241 in feces (part 3 of 3).

Visual Comparison of Data to Estimated Gross Alpha MDAs in Feces

1 E-2

1 E-1

1 E+0

1 E+1

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA of 0.6 (95%CL on Mean)

50th Percentile of 0.4

Figure 5A-16. MDA data for gross alpha in feces (part 1 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 63 of 70

Annual Average MDAs for Gross Alpha in Feces

Calculated from Recovered Data Between 1945 and 1990

0

0.2

0.4

0.6

0.8

1

1.2

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-16. MDA data for gross alpha in feces (part 2 of 3).

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Gross Alpha in Feces

0

20

40

60

80

100

120

140

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-16. MDA data for gross alpha in feces (part 3 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 64 of 70

Visual Comparison of Data to Estimated Pu-239 MDA in Feces

1 E-3

1 E-2

1 E-1

1 E+0

1 E+1

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA of 0.4 (95%CL on Mean)

50th Percentile of 0.1

Figure 5A-17. MDA data for Pu-239 in feces (part 1 of 3).

Annual Average MDAs for Pu-239 in Feces

Calculated from Recovered Data Between 1945 and 1990

0

0.2

0.4

0.6

0.8

1

1.2

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-17. MDA data for Pu-239 in feces (part 2 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 65 of 70

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Pu-239 in Feces

0

5

10

15

20

25

30

35

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-17. MDA data for Pu-239 in feces (part 3 of 3).

Visual Comparison of Data to Estimated Th-232 MDAs in Feces

1 E-2

1 E-1

1 E+0

1 E+1

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990

Year

Activity (dpm/24 h sample)

Sample MDA

Est. MDA of 0.5 (95%CL on Mean)

50th Percentile of 0.3

Figure 5A-18. MDA data for Th-232 in feces (part 1 of 3).
Effective Date: 06/29/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-5 Page 66 of 70

Annual Average MDAs for Th-232 in Feces

Calculated from Recovered Data Between 1945 and 1990

0

0.2

0.4

0.6

0.8

1

1.2

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

MDA (dpm/24 h)

Figure 5A-18. MDA data for Th-232 in feces (part 2 of 3).

Number of Sample Results Recovered per Year 

Between 1945 and 1990 for Th-232 in Feces 

0

100

200

300

400

500

600

700

800

1945194619471948194919501951195219531954195519561957195819591960196119621963196419651966196719681969197019711972197319741975197619771978197919801981198219831984198519861987198819891990

Year

Number of Samples (N)

Figure 5A-18. MDA data for Th-232 in feces (part 3 of 3).
Effective Date: 05/30/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0012-5 Page 67 of 70

ATTACHMENT 5B

ASSESSMENT OF INTERNAL DOSE AT ORNL: 1944 TO 1947

LIST OF FIGURES

Figure 

5B-1 Plutonium in urine samples from 1945 ...................................................................................

Page

68

5B-2 Plutonium urine sample data through April 18, 1945 ..............................................................68

5B-3 Plutonium urine sample data for April 25-May 30, 1945 .........................................................69

5B-4 Alpha air monitoring results for 1944-1947.............................................................................69

Information about internal dose at ORNL from 1943 to 1947 is somewhat limited, with no bioassay 

results appearing in the DOE worker files. ORAUT (2005) currently extends back to 1944 for use as 

an overestimate at the site because tolerance levels are well documented and there is strong 

evidence (memoramda and monthly reports) that exposure was controlled based on the results of the 

samples. These memoranda include weekly reports on the airborne activity levels in various areas, 

as well as precautions taken or to be taken in the future. Examples of these reports include Bloss

(1944a) to Bloss (1944g). 

ORAUT (2005) generally provides an overestimate for a number of reasons. Exposure at the 

maximum permissible air concentration is assumed for 40 hr per week, 50 weeks per year. While 

exposure to concentrations greater than this is likely over short periods for some occupations, the 

overriding control factor was to limit the average exposure over longer time frames. In addition, the 

nuclide that results in the largest dose is assumed to comprise 100% of the intake; the list of possible 

nuclides includes some that are very unlikely to be present in significant quantities. During the early 

years of ORNL, several of the nuclides on the list were just being discovered or were yet to be 

discovered. The assumed nuclide might change from year to year to ensure that the largest possible 

dose is assigned in each year until the date of diagnosis.

Although not included in the individual files, there are limited urine sample results from ORNL in this 

period. About 150 results for urine plutonium content were found; all were submitted in 1945. It is 

postulated that these are not in the database because they precede the routine sampling program at 

the site; they were collected as part of a study to determine “the amount of plutonium, if any, that may 

be fixed in the bodies of any chemists at Clinton Laboratories” (Wirth, 1945a). Urine samples were 

collected weekly for about 6 months at both Clinton (the original name for ORNL) and the Met Lab in 

Chicago. The entire data set was fitted to a lognormal distribution (Figure 5B-1). 

The geometric mean of this distribution is somewhat large (1.7 pCi/1400 mL), and the fit to the results 

indicates that there are possibly two distributions. If this value is assumed to be due to a constant, 

chronic intake of Type M 239Pu beginning in early 1944, an airborne concentration of 8.1 × 10-11 μCi 

/mL is indicated. This is almost three times larger than the tolerance level of 3 × 10-11 μCi /mL in 

effect at the time. If Type S is assumed, the airborne concentration is expected to be 2 orders of 

magnitude larger than the tolerance level.

In a memorandum dated May 11, 1945, it was reported that the findings appeared to be above control 

levels, which resulted in a more rigid technique of collection and system of collecting samples being 

implemented on April 25 (Wirth 1945b) . Results in the first 2 weeks of this new collection system 

were not substantially different from previous results, but samples in the subsequent, final 4 weeks of 

the study showed little detectable activity. Two separate lognormal fits were performed on these data, 

before and after the implementation of stricter controls. Figures 5B-2 and 5B-3 show the outcomes. 
Effective Date: 05/30/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0012-5 Page 68 of 70

0.01

0.1

1

10

100

1000

-2 -1 0 1 2 3

Z Score

Pu in Urine (pCi/1400 mL)

Figure 5B-1. Plutonium in urine samples from 1945. The geometric mean = 1.7 

pCi/day, the GSD = 5.3, and r

2 = 0.79.

0.01

0.1

1

10

100

1000

-3 -2 -1 0 1 2 3

Z Score

Pu in Urine (pCi/1400 mL)

Figure 5B-2. Plutonium urine sample data through April 18, 1945. The geometric 

mean = 3.1 pCi/day, the GSD = 4.2, and r

2 = 0.82.

The geometric mean before April 25, 1945, is 3.1 pCi/day, while after this date it drops to 

0.54 pCi/day. The arithmetic average of the results from April 25 through the end of the study is about 

25% of the average during the previous weeks, and the average from the final four weeks is about 3% 

of the earlier weeks. These comparisons indicate that the concerns about the collection system 

before April 25 may have been justified.

In addition to the urine samples, 1483 alpha air-monitoring results were available for 1944 through 

1947 at ORNL. The two largest values, both many orders of magnitude larger than the tolerance level 
Effective Date: 05/30/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0012-5 Page 69 of 70

0.01

0.1

1

10

100

-1 0 1 2 3

Z Score

Pu in Urine (pCi/1400 mL)

Figure 5B-3. Plutonium urine sample data for April 25-May 30, 1945. The 

geometric mean = 0.54 pCi/day, the GSD = 6.0, and r

2 = 0.80.

in use at the time, were excluded from the analysis because one appeared to be a miscalculation 

(based on the recorded alpha activity) and the other was collected inside of a hot cell. Results were fit 

to a lognormal distribution (Figure 5B-4). The geometric mean is 5% of the tolerance level applied in 

ORAUT (2005), and the 95th percentile is equal to the tolerance level. Notes indicating that assault 

masks had been worn were present for just under 10% of the samples; many of the samples were 

collected inside of hoods.

1.0E-14

1.0E-13

1.0E-12

1.0E-11

1.0E-10

1.0E-09

1.0E-08

1.0E-07

1.0E-06

-2 -1 0 1 2 3 4

Z Score

Air Sample Result (μCi/mL)

Figure 5B-4. Alpha air-monitoring results for 1944-1947. The geometric mean = 

1.4 × 10-12 μCi/mL, the GSD = 6.35, and r

2 = 0.97.
Effective Date: 05/30/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0012-5 Page 70 of 70

Given this information, ORAUT (2005) appears to be a reasonable upper bound because the 95th 

percentile of the air-monitoring results coincides with the value applied in that document. Assault 

masks were used in many situations, and many of the results exceed the activity to which the 

individual would have been exposed because samples were collected inside of hoods. There is 

conservancy built into ORAUT (2005) by the inclusion of alpha emitters with large dose coefficients 

that were likely not present or present only in very small quantities. 

Because it is not known who wore the assault masks and where specific individuals worked, it is 

difficult to postulate a smaller airborne concentration. Urine sample results are somewhat ambiguous 

because activity was detected in them but they were likely contaminated during collection or analysis. 

As a best estimate for individuals likely to have a potential for exposure, ORAUT (2005) values will be 

used but the nuclides will be limited to those likely to be present in significant quantities (plutonium 

and uranium isotopes), thus eliminating unnecessary conservancy. Due to the extent and quality of 

the alpha air-monitoring program, this approach is considered to be applicable to beta/gammaemitting radionuclides at ORNL during this period.